DANDE: a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs

DANDE-a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems

Overview of current RFSP-code capabilities for CANDU core analysis

International Nuclear Information System (INIS)

Rouben, B.

1996-01-01

RFSP (Reactor Fuelling Simulation Program) is the major finite-reactor computer code in use at the Atomic Energy of Canada Limited for the design and analysis of CANDU reactor cores. An overview is given of the major computational capabilities available in RFSP. (author) 11 refs., 29 figs

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

CEDNBR: a computer code for transient thermal margin analysis of a reactor core

International Nuclear Information System (INIS)

Shesler, A.T.; Lehmann, C.R.

1976-09-01

The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

International Nuclear Information System (INIS)

Basehore, K.L.; Todreas, N.E.

1980-08-01

Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries

Adaption of the PARCS Code for Core Design Audit Analyses

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyong Chol; Lee, Young Jin; Uhm, Jae Beop; Kim, Hyunjik [Nuclear Safety Evaluation, Daejeon (Korea, Republic of); Jeong, Hun Young; Ahn, Seunghoon; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

The eigenvalue calculation also includes quasi-static core depletion analyses. PARCS has implemented variety of features and has been qualified as a regulatory audit code in conjunction with other NRC thermal-hydraulic codes such as TRACE or RELAP5. In this study, as an adaptation effort for audit applications, PARCS is applied for an audit analysis of a reload core design. The lattice physics code HELIOS is used for cross section generation. PARCS-HELIOS code system has been established as a core analysis tool. Calculation results have been compared on a wide spectrum of calculations such as power distribution, critical soluble boron concentration, and rod worth. A reasonable agreement between the audit calculation and the reference results has been found.

Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

International Nuclear Information System (INIS)

Baiocco, G.; Petruzzi, A.; Bznuni, S.; Kozlowski, T.

2017-01-01

Highlights: • The consistency between Helios and Serpent few-group cross sections is shown. • The PARCS model is validated against a Monte Carlo 3D model. • The fission and capture rates are compared. • The influence of the spacer grids on the axial power distribution is shown. - Abstract: Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement.

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1996-07-01

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

Application of startup/core management code system to YGN 3 startup testing

International Nuclear Information System (INIS)

Chi, Sung Goo; Hah, Yung Joon; Doo, Jin Yong; Kim, Dae Kyum

1995-01-01

YGN 3 is the first nuclear power plant in Korea to use the fixed incore detector system for startup testing and core management. The startup/core management code system was developed from existing ABB-C-E codes and applied for YGN 3 startup testing, especially for physics and CPC(Core Protection Calculator)/COLSS (Core Operating Limit Supervisory System) related testing. The startup/core management code system consists of startup codes which include the CEBASE, CECOR, CEFAST and CEDOPS, and startup data reduction codes which include FLOWRATE, COREPERF, CALMET, and VARTAV. These codes were implemented on an HP/Apollo model 9000 series 400 workstation at the YGN 3 site and successfully applied to startup testing and core management. The startup codes made a great contribution in upgrading the reliability of test results and reducing the test period by taking and analyzing core data automatically. The data reduction code saved the manpower and time for test data reduction and decreased the chance for error in the analysis. It is expected that this code system will make similar contributions for reducing the startup testing duration of YGN 4 and UCN3,4

The Preliminary GAMMA Code Thermal hydraulic Analysis for the Steady State of HTR-10 Initial Core

Energy Technology Data Exchange (ETDEWEB)

Jun, Ji Su; Lim, Hong Sik; Lee, Won Jae

2006-07-15

This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a

Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

International Nuclear Information System (INIS)

Pavlovitchev, A.M.

2000-01-01

The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

International Nuclear Information System (INIS)

Shamasundar, B.I.; Fehrenbach, M.E.

1981-05-01

The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations

CONSUL code package application for LMFR core calculations

Energy Technology Data Exchange (ETDEWEB)

Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

2008-07-01

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

Analysis of the SPERT III E-core experiment using the EUREKA-2 code

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1986-09-01

EUREKA-2, a coupled nuclear thermal hydrodynamic kinetic code, was adapted for the testing of models and methods. Code evaluations were made with the reactivity addition experiments of the SPERT III E-Core, a slightly enriched oxide core. The code was tested for non damaging power excursions including a wide range of initial operating conditions, such as cold-startup, hot-startup, hot-standby and operating-power initial conditions. Comparisons resulted in a good agreement within the experimental errors between calculated and experimental power, energy, reactivity and clad surface temperature. (author)

Analysis of the AP600 core makeup tank experiments using the NOTRUMP code

International Nuclear Information System (INIS)

Cunningham, J.C.; Haberstroh, R.C.; Hochreiter, L.E.; Jaroszewicz, J.

1995-01-01

The AP600 design utilizes passive methods to perform core and containment cooling functions for a postulated loss of coolant. The core makeup tank (CMT) is an important feature of the AP600 passive safety system. The NOTRUMP code has been compared to the 300-series core makeup tank experiments. It has been observed that the code will capture the correct thermal-hydraulic behavior observed in the experiments. The correlations used for wall film condensation and convective heat transfer to the heated CMT liquid appear to be appropriate for these applications. The code will predict the rapid condensation and mixing thermal-hydraulic behavior observed in the 300-series tests. The NOTRUMP predictions can be noding-dependent since the condensation is extremely dependent on the amount of cold CMT liquid that mixes with the incoming steam flow

Validation study of core analysis methods for full MOX BWR

International Nuclear Information System (INIS)

2013-01-01

JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

Validation study of core analysis methods for full MOX BWR

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

Thermohydraulic characteristics analysis of natural convective cooling mode on the steady state condition of upgraded JRR-3 core, using COOLOD-N code

International Nuclear Information System (INIS)

Kaminaga, Masanori; Watanabe, Shukichi; Ando, Hiroei; Sudo, Yukio; Ikawa, Hiromasa.

1987-03-01

This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)

CINETHICA - Core accident analysis code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

Energy Technology Data Exchange (ETDEWEB)

Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

2014-10-15

The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

Kinetic parameters evaluation of PWRs using static cell and core calculation codes

International Nuclear Information System (INIS)

Jahanbin, Ali; Malmir, Hessam

2012-01-01

Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.

Verification of in-core thermal and hydraulic analysis code FLOWNET/TRUMP for the high temperature engineering test reactor (HTTR) at JAERI

International Nuclear Information System (INIS)

Maruyama, Soh; Sudo, Yukio; Saito, Shinzo; Kiso, Yoshihiro; Hayakawa, Hitoshi

1989-01-01

The FLOWNET/TRUMP code consists of a flow network analysis code 'FLOWNET' for calculations of coolant flow distribution and coolant temperature distribution in the core with a thermal conduction analysis code 'TRUMP' for calculation of temperature distribution in solid structures. The verification of FLOWNET/TRUMP was made by the comparison of the analytical results with the results of steady state experiments by the HENDEL multichannel test rig, T1-M, which consisted of twelve simulated fuel rods heated electrically and eleven hexagonal graphite fuel blocks. The T1-M simulated the one fuel column in the core. The analytical results agreed well with the results of the experiment in which the HTTR operating conditions were simulated. (orig.)

In-core fuel management code package validation for BWRs

International Nuclear Information System (INIS)

1995-12-01

The main goal of the present CRP (Coordinated Research Programme) was to develop benchmarks which are appropriate to check and improve the fuel management computer code packages and their procedures. Therefore, benchmark specifications were established which included a set of realistic data for running in-core fuel management codes. Secondly, the results of measurements and/or operating data were also provided to verify and compare with these parameters as calculated by the in-core fuel management codes or code packages. For the BWR it was established that the Mexican Laguna Verde 1 BWR would serve as the model for providing data on the benchmark specifications. It was decided to provide results for the first 2 cycles of Unit 1 of the Laguna Verde reactor. The analyses of the above benchmarks are performed in two stages. In the first stage, the lattice parameters are generated as a function of burnup at different voids and with and without control rod. These lattice parameters form the input for 3-dimensional diffusion theory codes for over-all reactor analysis. The lattice calculations were performed using different methods, such as, Monte Carlo, 2-D integral transport theory methods. Supercell Model and transport-diffusion model with proper correction for burnable absorber. Thus the variety of results should provide adequate information for any institute or organization to develop competence to analyze In-core fuel management codes. 15 refs, figs and tabs

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Comparative analysis of unprotected loss-of-flow accidents for the 1.0 m EFR-LVC core using different computer codes

International Nuclear Information System (INIS)

Royl, P.; Frizonnet, J.M.; Moran, J.

1993-02-01

A comparative analysis of the unprotected loss of flow (ULOF) accident has been performed for the LVC core (Lower Void Core) of the European Fast Reactor EFR with the FRAX5B and FRAX5C codes from the AEA-T, the PHYSURAC code from CEA and the SAS4A REF92 code system developed jointly between KfK, CEA and PNC. The accident is triggered by the run down of the coolant pumps with failure to trip the reactor by the primary and/or secondary shutdown system. Only a limited amount of mitigating reactivity from the third shutdown line was considered so that the accident can progress into boiling and core disruption. This code outlines the important modelling differences and compares the different simulations. The discussion of the rather wide spectrum of calculated accident progressions identifies the generic differences, relates them to the applied models, and summarizes the key points that are responsible for the different progressions. A comparison of the consequence spectrum from all simulations indicates zero work energies for the majority of the calculations. All simulations show up the need for a continued accident analysis into the early and late transition phase

Highly parallel line-based image coding for many cores.

Science.gov (United States)

Peng, Xiulian; Xu, Jizheng; Zhou, You; Wu, Feng

2012-01-01

Computers are developing along with a new trend from the dual-core and quad-core processors to ones with tens or even hundreds of cores. Multimedia, as one of the most important applications in computers, has an urgent need to design parallel coding algorithms for compression. Taking intraframe/image coding as a start point, this paper proposes a pure line-by-line coding scheme (LBLC) to meet the need. In LBLC, an input image is processed line by line sequentially, and each line is divided into small fixed-length segments. The compression of all segments from prediction to entropy coding is completely independent and concurrent at many cores. Results on a general-purpose computer show that our scheme can get a 13.9 times speedup with 15 cores at the encoder and a 10.3 times speedup at the decoder. Ideally, such near-linear speeding relation with the number of cores can be kept for more than 100 cores. In addition to the high parallelism, the proposed scheme can perform comparatively or even better than the H.264 high profile above middle bit rates. At near-lossless coding, it outperforms H.264 more than 10 dB. At lossless coding, up to 14% bit-rate reduction is observed compared with H.264 lossless coding at the high 4:4:4 profile.

Development of Regulatory Audit Core Safety Code : COREDAX

Energy Technology Data Exchange (ETDEWEB)

Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2005-07-01

Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

The Analysis of Surrounding Structure Effect on the Core Degradation Progress with COMPASS Code

Energy Technology Data Exchange (ETDEWEB)

Bae, Jun Ho; Son, Dong Gun; Kim, Jong Tae; Park, Rae Jun; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

2015-05-15

In line with the importance of severe accident analysis after Fukushima accident, the development of integrated severe accident code has been launched by the collaboration of three institutes in Korea. KAERI is responsible to develop modules related to the in-vessel phenomena, while other institutes are to the containment and severe accident mitigation facility, respectively. In the first phase, the individual severe accident module has been developed and the construction of integrated analysis code is planned to perform in the second phase. The basic strategy is to extend the design basis analysis codes of SPACE and CAP, which are being validated in Korea for the severe accident analysis. In the first phase, KAERI has targeted to develop the framework of severe accident code, COMPASS (COre Meltdown Progression Accident Simulation Software), covering the severe accident progression in a vessel from a core heat-up to a vessel failure as a stand-alone fashion. In order to analyze the effect of surrounding structure, the melt progression has been compared between the central zone and the most outer zone under the condition of constant radial power peaking factor. Figure 2 and 3 shows the fuel element temperature and the clad mass at the central zone, respectively. Due to the axial power peaking factor, the axial node No.3 has the highest temperature, while the top and bottom nodes have the lowest temperature. When the clad temperature reaches to the Zr melting temperature (2129.15K), the Zr starts to melt. The axial node No.2 reaches to the fuel melting temperature about 5000 sec and the molten fuel relocates to the node No.1, which results to the blockage of flow area in node No.1. The blocked flow area becomes to open about 6100 sec due to the molten ZrO{sub 2} mass relocation to core support plate. Figure 4 and 5 shows the fuel element temperature and the clad mass at the most outer zone, respectively. It is shown that the fuel temperature increase more slowly

Improvement of JRR-4 core management code system

International Nuclear Information System (INIS)

Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

2000-01-01

In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

HTGR core seismic analysis using an array processor

International Nuclear Information System (INIS)

Shatoff, H.; Charman, C.M.

1983-01-01

A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

Statistical core design methodology using the VIPRE thermal-hydraulics code

International Nuclear Information System (INIS)

Lloyd, M.W.; Feltus, M.A.

1995-01-01

An improved statistical core design methodology for developing a computational departure from nucleate boiling ratio (DNBR) correlation has been developed and applied in order to analyze the nominal 1.3 DNBR limit on Westinghouse Pressurized Water Reactor (PWR) cores. This analysis, although limited in scope, found that the DNBR limit can be reduced from 1.3 to some lower value and be accurate within an adequate confidence level of 95%, for three particular FSAR operational transients: turbine trip, complete loss of flow, and inadvertent opening of a pressurizer relief valve. The VIPRE-01 thermal-hydraulics code, the SAS/STAT statistical package, and the EPRI/Columbia University DNBR experimental data base were used in this research to develop the Pennsylvania State Statistical Core Design Methodology (PSSCDM). The VIPRE code was used to perform the necessary sensitivity studies and generate the EPRI correlation-calculated DNBR predictions. The SAS package used for these EPRI DNBR correlation predictions from VIPRE as a data set to determine the best fit for the empirical model and to perform the statistical analysis. (author)

DIANA Code: Design and implementation of an analytic core calculus code by two group, two zone diffusion

International Nuclear Information System (INIS)

Mochi, Ignacio

2005-01-01

The principal parameters of nuclear reactors are determined in the conceptual design stage.For that purpose, it is necessary to have flexible calculation tools that represent the principal dependencies of such parameters.This capability is of critical importance in the design of innovative nuclear reactors.In order to have a proper tool that could assist the conceptual design of innovative nuclear reactors, we developed and implemented a neutronic core calculus code: DIANA (Diffusion Integral Analytic Neutron Analysis).To calculate the required parameters, this code generates its own cross sections using an analytic two group, two zones diffusion scheme based only on a minimal set of data (i.e. 2200 m/s and fission averaged microscopic cross sections, Wescott factors and Effective Resonance Integrals).Both to calculate cross sections and core parameters, DIANA takes into account heterogeneity effects that are included when it evaluates each zone.Among them lays the disadvantage factor of each energy group.DIANA was totally implemented through Object Oriented Programming using C++ language. This eases source code understanding and would allow a quick expansion of its capabilities if needed.The final product is a versatile and easy-to-use code that allows core calculations with a minimal amount of data.It also contains the required tools needed to perform many variational calculations such as the parameterisation of effective multiplication factors for different radii of the core.The diffusion scheme s simplicity allows an easy following of the involved phenomena, making DIANA the most suitable tool to design reactors whose physics lays beyond the parameters of present reactors.All this reasons make DIANA a good candidate for future innovative reactor analysis

HTR core physics and transient analyses by the Panthermix code system

International Nuclear Information System (INIS)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

2005-01-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

International Nuclear Information System (INIS)

Hall, P.; Hutt, P.

1994-01-01

This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

International Nuclear Information System (INIS)

Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

1983-04-01

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

International Nuclear Information System (INIS)

Apostolov, T.; Ivanov, K.; Prodanova, R.; Manolova, M.; Petrova, T.; Alekova, G.

1993-01-01

Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

HTR core physics and transient analyses by the Panthermix code system

Energy Technology Data Exchange (ETDEWEB)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE

International Nuclear Information System (INIS)

Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa

2009-01-01

The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)

An analysis of the CSNI/GREST core concrete interaction chemical thermodynamic benchmark exercise using the MPEC2 computer code

International Nuclear Information System (INIS)

Muramatsu, Ken; Kondo, Yasuhiko; Uchida, Masaaki; Soda, Kunihisa

1989-01-01

Fission product (EP) release during a core concrete interaction (CCI) is an important factor of the uncertainty associated with a source term estimation for an LWR severe accident. An analysis was made on the CCI Chemical Thermodynamic Benchmark Exercise organized by OECD/NEA/CSNI Group of Experts on Source Terms (GREST) for investigating the uncertainty in thermodynamic modeling for CCI. The benchmark exercise was to calculate the equilibrium FP vapor pressure for given system of temperature, pressure, and debris composition. The benchmark consisted of two parts, A and B. Part A was a simplified problem intended to test the numerical techniques. In part B, the participants were requested to use their own best estimate thermodynamic data base to examine the variability of the results due to the difference in thermodynamic data base. JAERI participated in this benchmark exercise with use of the MPEC2 code. Chemical thermodynamic data base needed for analysis of Part B was taken from the VENESA code. This report describes the computer code used, inputs to the code, and results from the calculation by JAERI. The present calculation indicates that the FP vapor pressure depends strongly on temperature and Oxygen potential in core debris and the pattern of dependency may be different for different FP elements. (author)

Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

International Nuclear Information System (INIS)

Dozaki, Koji

2007-01-01

Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

A review of MAAP4 code structure and core T/H model

International Nuclear Information System (INIS)

Song, Yong Mann; Park, Soo Yong

1998-03-01

The modular accident analysis program (MAAP) version 4 is a computer code that can simulate the response of LWR plants during severe accident sequences and includes models for all of the important phenomena which might occur during accident sequences. In this report, MAAP4 code structure and core thermal hydraulic (T/H) model which models the T/H behavior of the reactor core and the response of core components during all accident phases involving degraded cores are specifically reviewed and then reorganized. This reorganization is performed via getting the related models together under each topic whose contents and order are same with other two reports for MELCOR and SCDAP/RELAP5 to be simultaneously published. Major purpose of the report is to provide information about the characteristics of MAAP4 core T/H models for an integrated severe accident computer code development being performed under the one of on-going mid/long-term nuclear developing project. The basic characteristics of the new integrated severe accident code includes: 1) Flexible simulation capability of primary side, secondary side, and the containment under severe accident conditions, 2) Detailed plant simulation, 3) Convenient user-interfaces, 4) Highly modularization for easy maintenance/improvement, and 5) State-of-the-art model selection. In conclusion, MAAP4 code has appeared to be superior for 3) and 4) items but to be somewhat inferior for 1) and 2) items. For item 5), more efforts should be made in the future to compare separated models in detail with not only other codes but also recent world-wide work. (author). 17 refs., 1 tab., 12 figs

An approach for coupled-code multiphysics core simulations from a common input

International Nuclear Information System (INIS)

Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott

2015-01-01

Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak

SACI - O: A code for the analysis of transients in a pressurized water reactor core

International Nuclear Information System (INIS)

Resende Lobo, A.A. de; Soares, P.A.

1979-03-01

The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt

ARKAS: A three-dimensional finite element code for the analysis of core distortions and mechanical behaviour

International Nuclear Information System (INIS)

Nakagawa, M.

1984-01-01

Computer program ARKAS has been developed for the purpose of predicting core distortions and mechanical behaviour in a cluster of subassemblies under steady state conditions in LMFBR cores. This report describes the analytical models and numerical procedures employed in the code together with some typical results of the analysis made on large LMFBR cores. ARKAS is programmed in the FORTRAN-IV language and is capable of treating up to 260 assemblies in a cluster with flexible boundary conditions including mirror and rotational symmetry. The nonlinearity of the problem due to contact and separation is solved by the step iterative procedure based on the Newton-Raphson method. In each step iterative procedure, the linear matrix equation must be reconstructed and then solved directly. To save computer time and memory, the substructure method is adopted in the step of reconstructing the linear matrix equation, and in the step of solving the linear matrix equation, the block successive over-relaxation method is adopted. The program ARKAS computes, at every time step, 3-dimensional displacements and rotations of the subassemblies in the core and the interduct forces including at the nozzle tips and nozzle bases with friction effects. The code also has an ability to deal with the refueling and shuffling of subassemblies and to calculate the values of withdrawal forces. For the qualitative validation of the code, sample calculations were performed on the several bundle arrays. In these calculations, contact and separation processes under the influences of friction forces, off-center loading, duct rotations and torsion, thermal expansion and irradiation induced swelling and creep were analyzed. These results are quite reasonable in the light of the expected behaviour. This work was performed under the sponsorship of Toshiba Corporation

Refuelling design and core calculations at NPP Paks: codes and methods

International Nuclear Information System (INIS)

Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

2001-01-01

This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

Qualification of ARROTTA code for LWR accident analysis

International Nuclear Information System (INIS)

Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

2004-01-01

This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

Preliminary evaluation of SACI-O code for the analysis of transients in a pressurized water reactor core

International Nuclear Information System (INIS)

Soares, P.A.; Sirimarco, L.F.; Veloso, M.A.F.

1979-03-01

SACI-O is a computer code which deals with the dynamics of the core of pressurized light water reactors (PWR). Its applicability is determined by the evaluation of the models used in the simulation of the several phenomena and processes which occur in the core during transients. This report presents a comparison between the results obtained with SACI-O and those presented in the Final Safety Analysis Report (FSAR) of Angra dos Reis Nuclear Station, Unit 1. Although some data used in the calculations done by Westinghouse are not known, there was a good agreement between the mentioned results. (Author) [pt

Parallelization of a three-dimensional whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Jin Young, Cho; Han Gyu, Joo; Ha Yong, Kim; Moon-Hee, Chang [Korea Atomic Energy Research Institute, Yuseong-gu, Daejon (Korea, Republic of)

2003-07-01

Parallelization of the DeCART (deterministic core analysis based on ray tracing) code is presented that reduces the computational burden of the tremendous computing time and memory required in three-dimensional whole core transport calculations. The parallelization employs the concept of MPI grouping and the MPI/OpenMP mixed scheme as well. Since most of the computing time and memory are used in MOC (method of characteristics) and the multi-group CMFD (coarse mesh finite difference) calculation in DeCART, variables and subroutines related to these two modules are the primary targets for parallelization. Specifically, the ray tracing module was parallelized using a planar domain decomposition scheme and an angular domain decomposition scheme. The parallel performance of the DeCART code is evaluated by solving a rodded variation of the C5G7MOX three dimensional benchmark problem and a simplified three-dimensional SMART PWR core problem. In C5G7MOX problem with 24 CPUs, a speedup of maximum 21 is obtained on an IBM Regatta machine and 22 on a LINUX Cluster in the MOC kernel, which indicates good parallel performance of the DeCART code. In the simplified SMART problem, the memory requirement of about 11 GBytes in the single processor cases reduces to 940 Mbytes with 24 processors, which means that the DeCART code can now solve large core problems with affordable LINUX clusters. (authors)

VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

International Nuclear Information System (INIS)

Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

1983-05-01

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

Implementation of refined core thermal-hydraulic calculation feature in the MARS/MASTER code

International Nuclear Information System (INIS)

Joo, H. K.; Jung, J. J.; Cho, B. O.; Ji, S. K.; Lee, W. J.; Jang, M. H.

2000-01-01

As an effort to enhance the fidelity of the core thermal/hydraulic calculation in the MARS/MASTER code, a best-estimate system/core coupled code, the COBRA-III module of MASTER is activated that enables refined core T/H calculations. Since the COBRA-III module is capable of using fuel-assembly sized nodes, the resolution of the T/H solution is high so that accurate incorporation of local T/H feedback effects becomes possible. The COBRA-III module is utilized such that the refined core T/H calculation is performed using the coarse-mesh flow boundary conditions specified by MARS at both ends of the core. The results of application to the OECD MSLB benchmark analysis indicate that the local peaking factor can be reduced by upto 15% with the refined calculation through the accurate representation of the local Doppler effect evaluation, although the prediction of the global transient behaviors such as the total core power change remain essentially unaffected

The UK core performance code package

International Nuclear Information System (INIS)

Hutt, P.K.; Gaines, N.; McEllin, M.; White, R.J.; Halsall, M.J.

1991-01-01

Over the last few years work has been co-ordinated by Nuclear Electric, originally part of the Central Electricity Generating Board, with contributions from the United Kingdom Atomic Energy Authority and British Nuclear Fuels Limited, to produce a generic, easy-to-use and integrated package of core performance codes able to perform a comprehensive range of calculations for fuel cycle design, safety analysis and on-line operational support for Light Water Reactor and Advanced Gas Cooled Reactor plant. The package consists of modern rationalized generic codes for lattice physics (WIMS), whole reactor calculations (PANTHER), thermal hydraulics (VIPRE) and fuel performance (ENIGMA). These codes, written in FORTRAN77, are highly portable and new developments have followed modern quality assurance standards. These codes can all be run ''stand-alone'' but they are also being integrated within a new UNIX-based interactive system called the Reactor Physics Workbench (RPW). The RPW provides an interactive user interface and a sophisticated data management system. It offers quality assurance features to the user and has facilities for defining complex calculational sequences. The Paper reviews the current capabilities of these components, their integration within the package and outlines future developments underway. Finally, the Paper describes the development of an on-line version of this package which is now being commissioned on UK AGR stations. (author)

Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

2007-07-15

Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

1991-05-01

The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

Analysis of Beryllium Having Irradiated at the RSG-GAS Core using ORIGEN2 Code

International Nuclear Information System (INIS)

Jaja Sukmana; Jonnie AK; S-Suwarto; Irwan

2012-01-01

Analysis of activation products generated by irradiated beryllium at the RSG-GAS core has been done using ORIGEN2 code. By assuming that irradiation is 176 days, neutron flux average of 2.30e+14 n/cm 2 s, radioisotopes rose from activated Be are tritium, lithium, beryllium, carbon, magnesium, aluminum, silicon, argon, calcium, scandium, vanadium, chromium, manganese, iron, cobalt, nickel, copper, zinc, silver, and lead. The highest activity after 100 days of irradiation demonstrated by Be-10 (7.99 E-03 Curie), H-3 (2.97 E-03 Curie), Cr-51, Fe-55 and Co-60. Radioactivity generated getting smaller when irradiation time are long. From this analysis it can be conclude that radioactivity was caused by impurities present in Be such as Mn-54, Fe-59, Zn-65, and Li-6. (author)

About the application of MCNP4 code in nuclear reactor core design calculations

International Nuclear Information System (INIS)

Svarny, J.

2000-01-01

This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

Systemization of burnup sensitivity analysis code

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2004-02-01

To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores

International Nuclear Information System (INIS)

Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.

1995-01-01

FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs

Preliminary analysis of the proposed BN-600 benchmark core

International Nuclear Information System (INIS)

John, T.M.

2000-01-01

The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

International Nuclear Information System (INIS)

Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao

2000-01-01

Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

Calculation of the RSG-GAS core using computer code citation-3D

International Nuclear Information System (INIS)

Taryo, T.; Rokhmadi

1998-01-01

Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that K eff are less and greater than unity (K eff eff >1)

Development of a BWR core burn-up calculation code COREBN-BWR

International Nuclear Information System (INIS)

Morimoto, Yuichi; Okumura, Keisuke

1992-05-01

In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

An analysis of the uniform core experiment

Energy Technology Data Exchange (ETDEWEB)

Waterson, R H

1973-10-15

This report describes an analysis of the Uniform Core of HITREX using the WIMS E codes, and presents the results of theory/experiment comparisons. The overall picture is one of good agreement for core reaction rate distributions, but theory umderestimating k{sub eff} by about 1.5% {delta}k/k.

KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

International Nuclear Information System (INIS)

Kim, Young Gyun; Kim, Young Il

2006-12-01

Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

International Nuclear Information System (INIS)

Gehre, G.

1982-01-01

First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

Benchmarking and qualification of the NUFREQ-NPW code for best estimate prediction of multi-channel core stability margins

International Nuclear Information System (INIS)

Taleyarkhan, R.; Lahey, R.T. Jr.; McFarlane, A.F.; Podowski, M.Z.

1988-01-01

The NUFREQ-NPW code was modified and set up at Westinghouse, USA for mixed fuel type multi-channel core-wide stability analysis. The resulting code, NUFREQ-NPW, allows for variable axial power profiles between channel groups and can handle mixed fuel types. Various models incorporated into NUFREQ-NPW were systematically compared against the Westinghouse channel stability analysis code MAZDA-NF, for which the mathematical model was developed, in an entirely different manner. Excellent agreement was obtained which verified the thermal-hydraulic modeling and coding aspects. Detailed comparisons were also performed against nuclear-coupled reactor core stability data. All thirteen Peach Bottom-2 EOC-2/3 low flow stability tests were simulated. A key aspect for code qualification involved the development of a physically based empirical algorithm to correct for the effect of core inlet flow development on subcooled boiling. Various other modeling assumptions were tested and sensitivity studies performed. Good agreement was obtained between NUFREQ-NPW predictions and data. Moreover, predictions were generally on the conservative side. The results of detailed direct comparisons with experimental data using the NUFREQ-NPW code; have demonstrated that BWR core stability margins are conservatively predicted, and all data trends are captured with good accuracy. The methodology is thus suitable for BWR design and licensing purposes. 11 refs., 12 figs., 2 tabs

Methodology for reactor core physics analysis - part 2

International Nuclear Information System (INIS)

Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

1992-12-01

The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

International Nuclear Information System (INIS)

Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

2002-01-01

This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

The PARET code and the analysis of the SPERT I transients

Energy Technology Data Exchange (ETDEWEB)

Woodruff, William L [Argonne National Laboratory, Argonne (United States)

1983-09-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

The PARET code and the analysis of the SPERT I transients

International Nuclear Information System (INIS)

Woodruff, William L.

1983-01-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

ELCOS: the PSI code system for LWR core analysis. Part II: user's manual for the fuel assembly code BOXER

International Nuclear Information System (INIS)

Paratte, J.M.; Grimm, P.; Hollard, J.M.

1996-02-01

ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user's manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs

Assessment of CANDU physics codes using experimental data - II: CANDU core physics measurements

International Nuclear Information System (INIS)

Roh, Gyu Hong; Jeong, Chang Joon; Choi, Hang Bok

2001-11-01

Benchmark calculations of the advanced CANDU reactor analysis tools (WIMS-AECL, SHETAN and RFSP) and the Monte Carlo code MCNP-4B have been performed using Wolsong Units 2 and 3 Phase-B measurement data. In this study, the benchmark calculations have been done for the criticality, boron worth, reactivity device worth, reactivity coefficient, and flux scan. For the validation of the WIMS-AECL/SHETANRFSP code system, the lattice parameters of the fuel channel were generated by the WIMS-AECL code, and incremental cross sections of reactivity devices and structural material were generated by the SHETAN code. The results have shown that the criticality is under-predicted by -4 mk. The reactivity device worths are generally consistent with the measured data except for the strong absorbers such as shutoff rod and mechanical control absorber. The heat transport system temperature coefficient and flux distributions are in good agreement with the measured data. However, the moderator temperature coefficient has shown a relatively large error, which could be caused by the incremental cross-section generation methodology for the reactivity device. For the MCNP-4B benchmark calculation, cross section libraries were newly generated from ENDF/B-VI release 3 through the NJOY97.114 data processing system and a three-dimensional full core model was developed. The simulation results have shown that the criticality is estimated within 4 mk and the estimated reactivity worth of the control devices are generally consistent with the measurement data, which implies that the MCNP code is valid for CANDU core analysis. In the future, therefore, the MCNP code could be used as a reference tool to benchmark design and analysis codes for the advanced fuels for which experimental data are not available

Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

International Nuclear Information System (INIS)

Mur, J.; Meignin, J.C.

1997-07-01

Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.)

Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

Energy Technology Data Exchange (ETDEWEB)

Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

1997-07-01

Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.

Applications of the thermit code to 3D thermal hydraulic analysis of LWR cores

International Nuclear Information System (INIS)

Reed, W.H.

1979-01-01

The THERMIT code calculates the three-dimensional transient thermal hydraulic behavior of light water reactor cores. Its two-fluid dynamics equations for two-phase flow offer improved physical modelling capability needed in the context of calculation coupled to neutron kinetics for feedback. The numerical fluid dynamics method was chosen for reliability over a wider range of transients. An improved heat transfer numerical method is presented which gives better numerical stability and accuracy. A number of example calculations are discussed which give an idea of the power and flexibility of the code

VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

International Nuclear Information System (INIS)

Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

1985-07-01

The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

CORA. A thermal and hydraulic transient analysis computer code for a cluster of reactor core assemblies

International Nuclear Information System (INIS)

Johnson, H.G.

1982-01-01

The Fast Flux Test Facility (FFTF) is arranged for natural circulation emergency core cooling in the event of loss of all plant electrical power. This design feature was conclusively demonstrated in a series of four natural circulation transient tests during the plant startup testing program in 1980 and 1981. Predictions, of core performance during these tests were made using the Westinghouse Hanford Company CORA computer program. The predictions, which compared well with measured plant data, were used in the extrapolation process to demonstrate the validity of the FFTF plant safety models and codes. This paper provides a brief description of the CORA code and includes typical comparisons of predictions to measured plant test data

Fuel management and core design code systems for pressurized water reactor neutronic calculations

International Nuclear Information System (INIS)

Ahnert, C.; Arayones, J.M.

1985-01-01

A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

International Nuclear Information System (INIS)

Mur, J.; Larrauri, D.

1998-07-01

Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

Integrated severe accident containment analysis with the CONTAIN computer code

International Nuclear Information System (INIS)

Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

1985-12-01

Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

International Nuclear Information System (INIS)

Lim, Hong Sik; Kim, Min Hwan

2009-01-01

The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

International Nuclear Information System (INIS)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

Development and application of sub-channel analysis code based on SCWR core

International Nuclear Information System (INIS)

Fu Shengwei; Xu Zhihong; Yang Yanhua

2011-01-01

The sub-channel analysis code SABER was developed for thermal-hydraulic analysis of supercritical water-cooled reactor (SCWR) fuel assembly. The extended computational cell structure, a new boundary conditions, 3 dimensional heat conduction model and water properties package were implemented in SABER code, which could be used to simulate the thermal fuel assembly of SCWR. To evaluate the applicability of the code, a steady state calculation of the fuel assembly was performed. The results indicate good applicability of the SABER code to simulate the counter-current flow and the heat exchange between coolant and moderator channels. (authors)

Improved core-edge tokamak transport simulations with the CORSICA 2 code

International Nuclear Information System (INIS)

Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

1996-01-01

The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

Analysis of the KUCA MEU experiments using the ANL code system

Energy Technology Data Exchange (ETDEWEB)

Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

1982-01-01

This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.

Analysis of the KUCA MEU experiments using the ANL code system

International Nuclear Information System (INIS)

Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

1982-01-01

This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis

Parallelization characteristics of a three-dimensional whole-core code DeCART

International Nuclear Information System (INIS)

Cho, J. Y.; Joo, H.K.; Kim, H. Y.; Lee, J. C.; Jang, M. H.

2003-01-01

Neutron transport calculation for three-dimensional amount of computing time but also huge memory. Therefore, whole-core codes such as DeCART need both also parallel computation and distributed memory capabilities. This paper is to implement such parallel capabilities based on MPI grouping and memory distribution on the DeCART code, and then to evaluate the performance by solving the C5G7 three-dimensional benchmark and a simplified three-dimensional SMART core problem. In C5G7 problem with 24 CPUs, a speedup of maximum 22 is obtained on IBM regatta machine and 21 on a LINUX cluster for the MOC kernel, which indicates good parallel performance of the DeCART code. The simplified SMART problem which need about 11 GBytes memory with one processors requires about 940 MBytes, which means that the DeCART code can now solve large core problems on affordable LINUX clusters

Scaling gysela code beyond 32K-cores on bluegene/Q***

Directory of Open Access Journals (Sweden)

Bigot J.

2013-12-01

Full Text Available Gyrokinetic simulations lead to huge computational needs. Up to now, the semi- Lagrangian code Gysela performed large simulations using a few thousands cores (8k cores typically. Simulation with finer resolutions and with kinetic electrons are expected to increase those needs by a huge factor, providing a good example of applications requiring Exascale machines. This paper presents our work to improve Gysela in order to target an architecture that presents one possible way towards Exascale: the Blue Gene/Q. After analyzing the limitations of the code on this architecture, we have implemented three kinds of improvement: computational performance improvements, memory consumption improvements and disk i/o improvements. As a result, we show that the code now scales beyond 32k cores with much improved performances. This will make it possible to target the most powerful machines available and thus handle much larger physical cases.

Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

International Nuclear Information System (INIS)

Ohtaka, Masahiko; Ohshima, Hiroyuki

1998-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

Development of core design/analysis technology for integral reactor; verification of SMART nuclear design by Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)

2002-03-01

The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)

ELCOS: the PSI code system for LWR core analysis. Part II: user`s manual for the fuel assembly code BOXER

Energy Technology Data Exchange (ETDEWEB)

Paratte, J.M.; Grimm, P.; Hollard, J.M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-02-01

ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user`s manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs.

Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

International Nuclear Information System (INIS)

Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

1997-01-01

The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

Development of an advanced code system for fast-reactor transient analysis

International Nuclear Information System (INIS)

Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

2005-01-01

FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

Multi-Core Processor Memory Contention Benchmark Analysis Case Study

Science.gov (United States)

Simon, Tyler; McGalliard, James

2009-01-01

Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

PC-Reactor-core transient simulation code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt

Systemization of burnup sensitivity analysis code. 2

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2005-02-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

Network Coding on Heterogeneous Multi-Core Processors for Wireless Sensor Networks

Science.gov (United States)

Kim, Deokho; Park, Karam; Ro, Won W.

2011-01-01

While network coding is well known for its efficiency and usefulness in wireless sensor networks, the excessive costs associated with decoding computation and complexity still hinder its adoption into practical use. On the other hand, high-performance microprocessors with heterogeneous multi-cores would be used as processing nodes of the wireless sensor networks in the near future. To this end, this paper introduces an efficient network coding algorithm developed for the heterogenous multi-core processors. The proposed idea is fully tested on one of the currently available heterogeneous multi-core processors referred to as the Cell Broadband Engine. PMID:22164053

Development of realistic thermal hydraulic system analysis code

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

Development of realistic thermal hydraulic system analysis code

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, B. D; Kim, K. D.

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

PanCoreGen - Profiling, detecting, annotating protein-coding genes in microbial genomes.

Science.gov (United States)

Paul, Sandip; Bhardwaj, Archana; Bag, Sumit K; Sokurenko, Evgeni V; Chattopadhyay, Sujay

2015-12-01

A large amount of genomic data, especially from multiple isolates of a single species, has opened new vistas for microbial genomics analysis. Analyzing the pan-genome (i.e. the sum of genetic repertoire) of microbial species is crucial in understanding the dynamics of molecular evolution, where virulence evolution is of major interest. Here we present PanCoreGen - a standalone application for pan- and core-genomic profiling of microbial protein-coding genes. PanCoreGen overcomes key limitations of the existing pan-genomic analysis tools, and develops an integrated annotation-structure for a species-specific pan-genomic profile. It provides important new features for annotating draft genomes/contigs and detecting unidentified genes in annotated genomes. It also generates user-defined group-specific datasets within the pan-genome. Interestingly, analyzing an example-set of Salmonella genomes, we detect potential footprints of adaptive convergence of horizontally transferred genes in two human-restricted pathogenic serovars - Typhi and Paratyphi A. Overall, PanCoreGen represents a state-of-the-art tool for microbial phylogenomics and pathogenomics study. Copyright © 2015 Elsevier Inc. All rights reserved.

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

Accelerator-driven transmutation reactor analysis code system (ATRAS)

Energy Technology Data Exchange (ETDEWEB)

Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-03-01

JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

Finite element program ARKAS: verification for IAEA benchmark problem analysis on core-wide mechanical analysis of LMFBR cores

International Nuclear Information System (INIS)

Nakagawa, M.; Tsuboi, Y.

1990-01-01

''ARKAS'' code verification, with the problems set in the International Working Group on Fast Reactors (IWGFR) Coordinated Research Programme (CRP) on the inter-comparison between liquid metal cooled fast breeder reactor (LMFBR) Core Mechanics Codes, is discussed. The CRP was co-ordinated by the IWGFR around problems set by Dr. R.G. Anderson (UKAEA) and arose from the IWGFR specialists' meeting on The Predictions and Experience of Core Distortion Behaviour (ref. 2). The problems for the verification (''code against code'') and validation (''code against experiment'') were set and calculated by eleven core mechanics codes from nine countries. All the problems have been completed and were solved with the core structural mechanics code ARKAS. Predictions by ARKAS agreed very well with other solutions for the well-defined verification problems. For the validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreements between measured and calculated values were fairly good. This paper briefly describes the numerical model of the ARKAS code, and discusses some typical results. (author)

Development of a perturbation code, PERT-K, for hexagonal core geometry

Energy Technology Data Exchange (ETDEWEB)

Kim, Taek Kyum; Kim, Sang Ji; Song, Hoon; Kim, Young Il; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

A perturbation code for hexagonal core geometry has been developed based on Nodal Expansion Method. By using relevant output files of DIF3D code, it can calculate the reactivity changes caused by perturbation in composition or/and neutron cross section libraries. The accuracy of PERT-K code has been validated by calculating the reactivity changes due to fuel composition change, the sodium void coefficients, and the sample reactivity worths of BFS-73-1 critical experiments. In the case of 10% reduction in all fuel isotopics at a assembly located in the outer core, PERT-K computation agrees with the direct computation by DIF3D within 60 pcm. The sample reactivity worths of BFS-73-1 critical experiments are predicted with PERT-K code within the experimental error bounds. For 100% sodium void occurrence at the inner core, the maximum difference of reactivity changes between PERT-K and direct DIF3D computations is less than 40 pcm. On the other hand, the same sodium void condition at the outer core leads to a difference of reactivity change greater than 400 pcm. However, as sodium voiding becomes near zero value, the difference becomes less and rapidly falls within the acceptable bound, i.e. 40 pcm. (author). 11 refs., 9 figs., 6 tabs.

Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code

International Nuclear Information System (INIS)

Chu, H.Y.; Chang, Y.W.

1977-01-01

This paper describes a computer code, MICE (Multifield Implicit Continuous-fluid Eulerian Containment Code), which is being developed at Argonne National Laboratory (ANL) for the analysis of containment response to a hypothetical core distruptive accident (HCDA). The code is applicable to multifield flow problems where material fields are allowed to have penetrations. Reactor structures are treated as axisymmetrical shells and solved by the large-displacement and small-strain theory. Two sample problems have been performed using the MICE code. The first illustrates the relative motions of the material fields after the initiation of a core disassembly accident. Core support structure and core barrel are modelled as rigid obstacles. The second demonstrates the interactions between fluid and structures. Core expansion and reactor wall deformation at several instants are shown by the computer-generated film plots. (Auth.)

Reactivity analysis of core distortion effects in the FFTF

International Nuclear Information System (INIS)

Knutson, B.J.

1982-01-01

An improved technique for evaluating core distortion reactivity effects was developed using reactivity analyses of two core geometry models (R-Z and HEX). This technique is incorporated into a new processor code called CORDIS. The advantages of this technique over existing reactivity models are that is preserves core heterogeneity, provides a control rod insertion effect model, uses row-dependent axial shape functions, and provides a flexible and cost efficient core distortion reactivity analysis method

User's manual for seismic analysis code 'SONATINA-2V'

International Nuclear Information System (INIS)

Hanawa, Satoshi; Iyoku, Tatsuo

2001-08-01

The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

Improvement of molten core-concrete interaction model of the debris spreading analysis model in the SAMPSON code - 15193

International Nuclear Information System (INIS)

Hidaka, M.; Fujii, T.; Sakai, T.

2015-01-01

A debris spreading analysis (DSA) module has been developed and improved. The module is used in the severe accident analysis code SAMPSON and it has models for 3-dimensional natural convection with simultaneous spreading, melting and solidification. The existing analysis method of the quasi-3D boundary transportation to simulate downward concrete erosion for evaluation of molten-core concrete interaction (MCCI) was improved to full-3D to solve, for instance, debris lateral erosion under concrete floors at the bottom of the sump pit. In the advanced MCCI model, buffer cells were defined in order to solve numerical problems in case of trammel formation. Mass, momentum, and the advection term of energy between the debris melt cells and the buffer cells are solved. On the other hand, only the heat transfer and thermal conduction are solved between the debris melt cells and the structure cells, and the crust cells and the structure cells. As a preliminary analysis, a validation calculation was performed for erosion that occurred in the core-concrete interaction (CCI-2) test in the OECD/MCCI program. Comparison between the calculation and the CCI-2 test results showed the analysis has the ability to simulate debris lateral erosion under concrete floors. (authors)

On-line generation of core monitoring power distribution in the SCOMS couppled with core design code

International Nuclear Information System (INIS)

Lee, K. B.; Kim, K. K.; In, W. K.; Ji, S. K.; Jang, M. H.

2002-01-01

The paper provides the description of the methodology and main program module of power distribution calculation of SCOMS(SMART COre Monitoring System). The simulation results of the SMART core using the developed SCOMS are included. The planar radial peaking factor(Fxy) is relatively high in SMART core because control banks are inserted to the core at normal operation. If the conventional core monitoring method is adapted to SMART, highly skewed planar radial peaking factor Fxy yields an excessive conservatism and reduces the operation margin. In addition to this, the error of the core monitoring would be enlarged and thus operating margin would be degraded, because it is impossible to precalculate the core monitoring constants for all the control banks configurations taking into account the operation history in the design stage. To get rid of these drawbacks in the conventional power distribution calculation methodology, new methodology to calculate the three dimensional power distribution is developed. Core monitoring constants are calculated with the core design code (MASTER) which is on-line coupled with SCOMS. Three dimensional (3D) power distribution and the several peaking factors are calculated using the in-core detector signals and core monitoring constant provided at real time. Developed methodology is applied to the SMART core and the various core states are simulated. Based on the simulation results, it is founded that the three dimensional peaking factor to calculate the Linear Power Density and the pseudo hot-pin axial power distribution to calculate the Departure Nucleate Boiling Ratio show the more conservative values than those of the best-estimated core design code, and SCOMS adapted developed methodology can secures the more operation margin than the conventional methodology

Analysis of the thermal hydraulics and core degradation behavior in the PHEBUS-FPT1 test train with impact/SAMPSON code

International Nuclear Information System (INIS)

Terada, Masafumi; Ikeda, Takashi; Nakahara, Katsuhiko; Shirakawa, Noriyuki; Horie, Hideki; Katsuragi, Kazuyuki; Yamagishi, Makoto; Ito, Takahiro

2003-01-01

As one of the verification studies of SAMPSON code, PHEBUS-FPT1, which is authorized as the International Standard Problem-46, was analyzed about the in-core phenomena with four modules, the molten core relocation analysis (MCRA) module, the fuel rod heat up analysis (FRHA) module, the fission product release analysis (FPRA) module, and the analysis control module (ACM) of SAMPSON. This paper describes the analysis of thermal hydraulics and core degradation behavior in the test train. Two-dimensional version of MCRA models the whole structure of the test train in the cylindrical system, including the fuel bundle and the shroud. FRHA models eighteen irradiated fuel rods, two fresh fuel rods, and one control rod in the center of the bundle. FRHA evaluates the transient behavior of fuel rods and releases failed fuel components to MCRA. MCRA evaluates the fluid dynamics of steam and debris considering the thermal and fluid mechanical interaction between them, and at the same time the thermal interaction between gas/debris and shroud material. By the phase change model of MCRA, molten debris forms debris pool and a part of them possibly freezes on fuel rods or shroud surface, then forms crust. This combination of modules of SAMPSON was proved to be capable for modeling the PHEBUS-FPT1 in-core phenomena sufficiently. The analysis has shown sufficient agreement with test results regarding to steam flow rates at the outlet, reproducing its reduction due to hydrogen generation, steam and shroud temperature, and debris relocation behavior. (author)

PWR core safety analysis with 3-dimensional methods

International Nuclear Information System (INIS)

Gensler, A.; Kühnel, K.; Kuch, S.

2015-01-01

Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

The in-core fuel management code system for VVER reactors

International Nuclear Information System (INIS)

Cada, R.; Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.

2004-01-01

The structure and methodology of a fuel management system for NPP VVER 1000 (NPP Temelin) and VVER 440 (NPP Dukovany) is described. It is under development in SKODA JS a.s. and is followed by practical applications. The general objectives of the system are maximization of end of cycle reactivity, the minimization of fresh fuel inventory for the minimization of fed enrichment and minimization of burnable poisons (BPs) inventory. They are also safety related constraints in witch minimization of power peaking plays a dominant role. General structure of the system consists in preparation of input data for macrocode calculation, algorithms (codes) for optimization of fuel loading, calculation of fuel enrichment and BPs assignment. At present core loading can be calculated (optimized) by Tabu search algorithm (code ATHENA), genetic algorithm (code Gen1) and hybrid algorithm - simplex procedure with application of Tabu search algorithm on binary shuffling (code OPAL B ). Enrichment search is realized by the application of simplex algorithm (OPAL B code) and BPs assignment by module BPASS and simplex algorithm in OPAL B code. Calculations of the real core loadings are presented and a comparison of different optimization methods is provided. (author)

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

Energy Technology Data Exchange (ETDEWEB)

Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

International Nuclear Information System (INIS)

Kang, S. H.; Ha, K. S.

2013-01-01

The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity feedback (Doppler and

JAERI thermal reactor standard code system for reactor design and analysis SRAC

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1985-01-01

SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

Core design calculations of IRIS reactor using modified CORD-2 code package

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

2002-01-01

Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

Integrated analysis of core debris interactions and their effects on containment integrity using the CONTAIN computer code

International Nuclear Information System (INIS)

Carroll, D.E.; Bergeron, K.D.; Williams, D.C.; Tills, J.L.; Valdez, G.D.

1987-01-01

The CONTAIN computer code includes a versatile system of phenomenological models for analyzing the physical, chemical and radiological conditions inside the containment building during severe reactor accidents. Important contributors to these conditions are the interactions which may occur between released corium and cavity concrete. The phenomena associated with interactions between ejected corium debris and the containment atmosphere (Direct Containment Heating or DCH) also pose a potential threat to containment integrity. In this paper, we describe recent enhancements of the CONTAIN code which allow an integrated analysis of these effects in the presence of other mitigating or aggravating physical processes. In particular, the recent inclusion of the CORCON and VANESA models is described and a calculation example presented. With this capability CONTAIN can model core-concrete interactions occurring simultaneously in multiple compartments and can couple the aerosols thereby generated to the mechanistic description of all atmospheric aerosol components. Also discussed are some recent results of modeling the phenomena involved in Direct Containment Heating. (orig.)

SASSYS LMFBR systems analysis code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.

1982-01-01

The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

HTR core physics analysis at NRG

International Nuclear Information System (INIS)

Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

2002-01-01

Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

International Nuclear Information System (INIS)

Dimov, D.

2011-01-01

The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

The verification of PWR-fuel code for PWR in-core fuel management

International Nuclear Information System (INIS)

Surian Pinem; Tagor M Sembiring; Tukiran

2015-01-01

In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

Analysis of neutronic parameters of AP1000 core for 18 month and 16/20 month cycle schemes using CASMO4E and SIMULATE-3 codes

International Nuclear Information System (INIS)

Nawaz Amjad; Yoshikawa, Hidekazu; Ming Yang

2015-01-01

AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC. (author)

Application of Looped Network Analysis Method to Core of Prismatic VHTR

International Nuclear Information System (INIS)

Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

2016-01-01

Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively as shown in Fig. 1. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Moreover, it is hard to cover whole cases corresponding to the various bypass gap distribution in the whole VHTR core. In order to solve this problem, in this study, the flow network analysis code, FastNet (Flow Analysis for Steady-state Network), was developed using the Looped Network Analysis Method. The applied method was validated by comparing with SNU VHTR multi-block experiment. A 3-demensional network modeling was conducted representing flow paths as flow resistances. Flow network analysis code, FastNet, was developed to evaluate the core bypass flow distribution by using looped network analysis method. Complex flow network could be solved simply by converting the non-linear momentum equation to the linearized equation. The FastNet code predicted the flow distribution of the SNU multi-block experiment accurately

Computer code HYDRO-ACE for analyzing thermo-hydraulic phenomena in the BWR core

International Nuclear Information System (INIS)

Abe, Kiyoharu; Naito, Yoshitaka

1979-10-01

A computer code HYDRO-ACE has been developed for analyzing thermo-hydraulic phenomena in the BWR core under forced or natural circulation of cooling water. The code is composed of two main calculation routines for single channels such as riser, separator, and downcommer and multiple channels such as the reactor core with a heated zone. Functionally the code is divided into many subroutines to be connected straightforwardly, and so that the user can choose a given course freely by simply arranging the subroutines. In the program, void fraction is calculated by Maurer's method, two-phase frictional pressure drop by Maltinelli-Nelson's, and critical heat flux ratio by Hench-Levy's. The coolant flow distributions in the JPDR-II core calculated by the code are in good agreement with those measured. (author)

Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

International Nuclear Information System (INIS)

1997-08-01

In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

Development of the SCHAMBETA code for scoping analysis of HCDA

Energy Technology Data Exchange (ETDEWEB)

Suk, Soo Dong; Hahn, D. H

2000-06-01

A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.

Analysis of space-time core dynamics on reactor accident at Chernobyl

International Nuclear Information System (INIS)

Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

1987-05-01

Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

A nodal Grean's function method of reactor core fuel management code, NGCFM2D

International Nuclear Information System (INIS)

Li Dongsheng; Yao Dong.

1987-01-01

This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

Transient and fuel performance analysis with VTT's coupled code system

International Nuclear Information System (INIS)

Daavittila, A.; Hamalainen, A.; Raty, H.

2005-01-01

VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

PanCoreGen – profiling, detecting, annotating protein-coding genes in microbial genomes

Science.gov (United States)

Bhardwaj, Archana; Bag, Sumit K; Sokurenko, Evgeni V.

2015-01-01

A large amount of genomic data, especially from multiple isolates of a single species, has opened new vistas for microbial genomics analysis. Analyzing pan-genome (i.e. the sum of genetic repertoire) of microbial species is crucial in understanding the dynamics of molecular evolution, where virulence evolution is of major interest. Here we present PanCoreGen – a standalone application for pan- and core-genomic profiling of microbial protein-coding genes. PanCoreGen overcomes key limitations of the existing pan-genomic analysis tools, and develops an integrated annotation-structure for species-specific pan-genomic profile. It provides important new features for annotating draft genomes/contigs and detecting unidentified genes in annotated genomes. It also generates user-defined group-specific datasets within the pan-genome. Interestingly, analyzing an example-set of Salmonella genomes, we detect potential footprints of adaptive convergence of horizontally transferred genes in two human-restricted pathogenic serovars – Typhi and Paratyphi A. Overall, PanCoreGen represents a state-of-the-art tool for microbial phylogenomics and pathogenomics study. PMID:26456591

Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

International Nuclear Information System (INIS)

Murata, Isao; Miyamaru, Hiroyuki

2008-01-01

Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%Δk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

Analysis of Random-Loading HTR-PROTEUS Cores with Continuous Energy Monte Carlo Code Based on A Statistical Geometry Model

Energy Technology Data Exchange (ETDEWEB)

Murata, Isao; Miyamaru, Hiroyuki [Division of Electrical, Electronic and Information Engineering, Osaka University, Yamada-oka 2-1, Suita, Osaka, 565-0871 (Japan)

2008-07-01

Spherical elements have remarkable features in various applications in the nuclear engineering field. In 1990's, by the project of HTR-PROTEUS at PSI various pebble bed reactor experiments were conducted including cores with a lot of spherical fuel elements loaded randomly. In this study, criticality experiments of the random-loading HTR-PROTEUS cores were analyzed by MCNP-BALL, which could deal with a random arrangement of spherical fuel elements exactly with a statistical geometry model. As a result of analysis, the calculated effective multiplication factors were in fairly good agreement with the measurements within about 0.5%DELTAk/k. In comparison with other numerical analysis, our effective multiplication factors were between the experimental values and the VSOP calculations. To investigate the discrepancy of the effective multiplication factors between the experiments and calculations, sensitivity analyses were performed. As the result, the sensitivity of impurity boron concentration was fairly large. The reason of the present slight overestimation was not made clear at present. However, the presently existing difference was thought to be related to the impurity boron concentration, not to the modelling of the reactor and the used nuclear data. From the present study, it was confirmed that MCNP-BALL would have an advantage to conventional transport codes by comparing with their numerical results and the experimental values. As for the criticality experiment of PROTEUS, we would conclude that the two cores of Core 4.2 and 4.3 could be regarded as an equivalent experiment of a reference critical core, which was packed in the packing fraction of RLP. (authors)

Safety analysis of RSG-GAS Silicide core using one line cooling system

International Nuclear Information System (INIS)

Endiah-Puji-Hastuti

2003-01-01

In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor has been determined and to continuing this program, steady state and transient analysis were done. The analysis was done by means of a core thermal hydraulic code, COOLOD-N, and PARET. The codes solves core thermal hydraulic equation at steady state conditions and transient, respectively. By using silicide core data and coast down flow rate as the input, thermal hydraulics parameters such as fuel cladding and fuel meat temperatures as well as safety margin against flow instability were calculated. Imposing the safety criteria to the results of steady state and transient analysis, maximum permissible power for this operation was obtained as much as 17.1 MW

The new lattice code Paragon and its qualification for PWR core applications

International Nuclear Information System (INIS)

Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

2003-01-01

Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

Energy Technology Data Exchange (ETDEWEB)

Wan, Chenghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Shen, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2017-04-15

Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k{sub eff} and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

Uncertainty analysis for the assembly and core simulation of BEAVRS at the HZP conditions

International Nuclear Information System (INIS)

Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

2017-01-01

Highlights: • Uncertainty analysis has been completed based on the “two-step” scheme. • Uncertainty analysis has been performed to BEAVRS at HZP. • For lattice calculations, the few-group constant’s uncertainty was quantified. • For core simulation, uncertainties of k_e_f_f and power distributions were quantified. - Abstract: Based on the “two-step” scheme for the reactor-physics calculations, the capability of uncertainty analysis for the core simulations has been implemented in the UNICORN code, an in-house code for the sensitivity and uncertainty analysis of the reactor-physics calculations. Applying the statistical sampling method, the nuclear-data uncertainties can be propagated to the important predictions of the core simulations. The uncertainties of the few-group constants introduced by the uncertainties of the multigroup microscopic cross sections are quantified first for the lattice calculations; the uncertainties of the few-group constants are then propagated to the core multiplication factor and core power distributions for the core simulations. Up to now, our in-house lattice code NECP-CACTI and the neutron-diffusion solver NECP-VIOLET have been implemented in UNICORN for the steady-state core simulations based on the “two-step” scheme. With NECP-CACTI and NECP-VIOLET, the modeling and simulation of the steady-state BEAVRS benchmark problem at the HZP conditions was performed, and the results were compared with those obtained by CASMO-4E. Based on the modeling and simulation, the UNICORN code has been applied to perform the uncertainty analysis for BAEVRS at HZP. The uncertainty results of the eigenvalues and two-group constants for the lattice calculations and the multiplication factor and the power distributions for the steady-state core simulations are obtained and analyzed in detail.

Operational reactor physics analysis codes (ORPAC)

International Nuclear Information System (INIS)

Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

2007-07-01

For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

Science.gov (United States)

Karim, Julia Abdul

2008-05-01

The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

International Nuclear Information System (INIS)

Karim, Julia Abdul

2008-01-01

The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained

Computer code validation study of PWR core design system, CASMO-3/MASTER-α

International Nuclear Information System (INIS)

Lee, K. H.; Kim, M. H.; Woo, S. W.

1999-01-01

In this paper, the feasibility of CASMO-3/MASTER-α nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-α was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-α calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-α was validated for commercial PWR core

Parallelization of Subchannel Analysis Code MATRA

International Nuclear Information System (INIS)

Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk

2014-01-01

A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems

A NEM diffusion code for fuel management and time average core calculation

International Nuclear Information System (INIS)

Mishra, Surendra; Ray, Sherly; Kumar, A.N.

2005-01-01

A computer code based on Nodal expansion method has been developed for solving two groups three dimensional diffusion equation. This code can be used for fuel management and time average core calculation. Explicit Xenon and fuel temperature estimation are also incorporated in this code. TAPP-4 phase-B physics experimental results were analyzed using this code and a code based on FD method. This paper gives the comparison of the observed data and the results obtained with this code and FD code. (author)

FAST: An advanced code system for fast reactor transient analysis

International Nuclear Information System (INIS)

Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

2005-01-01

One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

Analysis of high moderation full MOX BWR core physics experiments BASALA

International Nuclear Information System (INIS)

Ishii, Kazuya; Ando, Yoshihira; Takada, Naoyuki; Kan, Taro; Sasagawa, Masaru; Kikuchi, Tsukasa; Yamamoto, Toru; Kanda, Ryoji; Umano, Takuya

2005-01-01

Nuclear Power Engineering Corporation (NUPEC) has performed conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program BASALA following MISTRAL. They were devoted to measuring the core physics parameters of such advanced cores. The MISTRAL program consists of one reference UO 2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. As for MISTRAL, the analysis results have already been reported on April 2003. The BASALA program consists of two high moderation full MOX BWR mock-up cores for operating and cold stand-by conditions. NUPEC has analyzed the experimental results of BASALA with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. The analysis results of MISTRAL and BASALA indicate that these applied analysis methods have the same accuracy for the UO 2 and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores. (author)

CONTEMPT-DG containment analysis code

International Nuclear Information System (INIS)

Deem, R.E.; Rousseau, K.

1982-01-01

The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given

PWR core and spent fuel pool analysis using scale and nestle

International Nuclear Information System (INIS)

Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

2012-01-01

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

PWR core and spent fuel pool analysis using scale and nestle

Energy Technology Data Exchange (ETDEWEB)

Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

2012-07-01

The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

International Nuclear Information System (INIS)

Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

2002-01-01

A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

Conceptual study of advanced PWR core design

International Nuclear Information System (INIS)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

Conceptual study of advanced PWR core design

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

1997-09-01

The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

Monte carlo depletion analysis of SMART core by MCNAP code

International Nuclear Information System (INIS)

Jung, Jong Sung; Sim, Hyung Jin; Kim, Chang Hyo; Lee, Jung Chan; Ji, Sung Kyun

2001-01-01

Depletion an analysis of SMART, a small-sized advanced integral PWR under development by KAERI, is conducted using the Monte Carlo (MC) depletion analysis program, MCNAP. The results are compared with those of the CASMO-3/ MASTER nuclear analysis. The difference between MASTER and MCNAP on k eff prediction is observed about 600pcm at BOC, and becomes smaller as the core burnup increases. The maximum difference bet ween two predict ions on fuel assembly (FA) normalized power distribution is about 6.6% radially , and 14.5% axially but the differences are observed to lie within standard deviation of MC estimations

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

Energy Technology Data Exchange (ETDEWEB)

Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

2014-10-01

Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

International Nuclear Information System (INIS)

Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

2014-01-01

Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

User's manual of SECOM2: a computer code for seismic system reliability analysis

International Nuclear Information System (INIS)

Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Tamura, Kazuo

2002-03-01

This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, and Efficient sensitivity analysis by changing parameters on responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about response and capacity of components and seismic hazard curve for the NPP site as inputs. This report presents the models and methods applied in the SECOM2 code and how to use those functions. (author)

Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

International Nuclear Information System (INIS)

Hashimoto, Kazuichiro; Soda, Kunihisa

1991-10-01

The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

Electronic manual of the nuclear characteristics analysis code-set for FBR

International Nuclear Information System (INIS)

Makino, Tohru

2001-03-01

Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

Energy Technology Data Exchange (ETDEWEB)

Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

International Nuclear Information System (INIS)

Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

2003-01-01

The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

Evaluation of the WIMS (KAERI) - VENTURE code system for peak power prediction of KMRR core using MCNP

International Nuclear Information System (INIS)

Park, W.S.; Lee, K.M.; Lee, C.S.; Lee, J.T.; Oh, S.K.

1992-01-01

In this work, the validity and quantitative uncertainty of WIMS (KAERI) - VENTURE code system for the design and analysis of KMRR core was tried to be inferred using a well known benchmark code, MCNP. WIMS (KAERI) showed an excellent agreement with MCNP code. For three different control rod positions at a simulated core which has a quarter symmetry, total peaking factors and three sub-factors (radial, axial, and local) obtained from VENTURE were compared with those of MCNP. The comparison proved the validity of VENTURE and showed better agreement in the order of radial, axial, and local factors. The uncertainty of WIMS (KAERI) - VENTURE system was inferred using the 2σ band of total peaking obtained by MCNP. The uncertainty of WIMS (KAERI) - VENTURE system were found to be 18.5 % for the operating condition. (author)

Thermal hydraulic analysis of the JMTR improved LEU-core

Energy Technology Data Exchange (ETDEWEB)

Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

2003-01-01

After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

Modelling guidelines for core exit temperature simulations with system codes

Energy Technology Data Exchange (ETDEWEB)

Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

2015-05-15

Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

International Nuclear Information System (INIS)

Won-Seok Kim; Young-Gyun Kim

2000-01-01

In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

Benchmarking and qualification of the ppercase nufreq -ppercase npw code for best estimate prediction of multichannel core stability margins

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; McFarlane, A.F.; Lahey, R.T. Jr.; Podowski, M.Z.

1994-01-01

The ppercase nufreq - ppercase np (G.C. Park et al. NUREG/CR-3375, 1983; S.J. Peng et al. NUREG/CR-4116, 1984; S.J. Peng et al. Nucl. Sci. Eng. 88 (1988) 404-411) code was modified and set up at Westinghouse, USA, for mixed fuel type multichannel core-wide stability analysis. The resulting code, ppercase nufreq - ppercase npw , allows for variable axial power profiles between channel groups and can handle mixed fuel types.Various models incorporated into ppercase nurfreq - ppercase npw were systematically compared against the Westinghouse channel stability analysis code ppercase mazda -ppercase nf (R. Taleyarkhan et al. J. Heat Transfer 107 (February 1985) 175-181; NUREG/CR2972, 1983), for which the mathematical model was developed in an entirely different manner. Excellent agreement was obtained which verified the thermal-hydraulic modeling and coding aspects. Detailed comparisons were also performed against nuclear-coupled reactor core stability data. All 13 Peach Bottom-2 EOC-2/3 low flow stability tests (L.A. Carmichael and R.O. Neimi, EPRI NP-564, Project 1020-1, 1978; F.B. Woffinden and R.O. Neimi, EPRI, NP 0972, Project 1020-2, 1981) were simulated. A key aspect for code qualification involved the development of a physically based empirical algorithm to correct for the effect of core inlet flow development on subcooled boiling. Various other modeling assumptions were tested and sensitivity studies performed. Good agreement was obtained between ppercase nufreq-npw predictions and data. ((orig.))

Systemization of burnup sensitivity analysis code (2) (Contract research)

International Nuclear Information System (INIS)

Tatsumi, Masahiro; Hyoudou, Hideaki

2008-08-01

Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

Benchmarking and qualification of the nufreq-npw code for best estimate prediction of multi-channel core stability margins

International Nuclear Information System (INIS)

Taleyarkhan, R.; McFarlane, A.F.; Lahey, R.T. Jr.; Podowski, M.Z.

1988-01-01

The work described in this paper is focused on the development, verification and benchmarking of the NUFREQ-NPW code at Westinghouse, USA for best estimate prediction of multi-channel core stability margins in US BWRs. Various models incorporated into NUFREQ-NPW are systematically compared against the Westinghouse channel stability analysis code MAZDA, which the Mathematical Model was developed in an entirely different manner. The NUFREQ-NPW code is extensively benchmarked against experimental stability data with and without nuclear reactivity feedback. Detailed comparisons are next performed against nuclear-coupled core stability data. A physically based algorithm is developed to correct for the effect of flow development on subcooled boiling. Use of this algorithm (to be described in the full paper) captures the peak magnitude as well as the resonance frequency with good accuracy

Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

Energy Technology Data Exchange (ETDEWEB)

Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2016-09-15

A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

International Nuclear Information System (INIS)

Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

2000-01-01

This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

Engineering application of in-core fuel management optimization code with CSA algorithm

Energy Technology Data Exchange (ETDEWEB)

Liu, Zhihong; Hu, Yongming [INET, Tsinghua university, Beijing 100084 (China)

2009-06-15

PWR in-core loading (reloading) pattern optimization is a complex combined problem. An excellent fuel management optimization code can greatly improve the efficiency of core reloading design, and bring economic and safety benefits. Today many optimization codes with experiences or searching algorithms (such as SA, GA, ANN, ACO) have been developed, while how to improve their searching efficiency and engineering usability still needs further research. CSA (Characteristic Statistic Algorithm) is a global optimization algorithm with high efficiency developed by our team. The performance of CSA has been proved on many problems (such as Traveling Salesman Problems). The idea of CSA is to induce searching direction by the statistic distribution of characteristic values. This algorithm is quite suitable for fuel management optimization. Optimization code with CSA has been developed and was used on many core models. The research in this paper is to improve the engineering usability of CSA code according to all the actual engineering requirements. Many new improvements have been completed in this code, such as: 1. Considering the asymmetry of burn-up in one assembly, the rotation of each assembly is considered as new optimization variables in this code. 2. Worth of control rods must satisfy the given constraint, so some relative modifications are added into optimization code. 3. To deal with the combination of alternate cycles, multi-cycle optimization is considered in this code. 4. To confirm the accuracy of optimization results, many identifications of the physics calculation module in this code have been done, and the parameters of optimization schemes are checked by SCIENCE code. The improved optimization code with CSA has been used on Qinshan nuclear plant of China. The reloading of cycle 7, 8, 9 (12 months, no burnable poisons) and the 18 months equilibrium cycle (with burnable poisons) reloading are optimized. At last, many optimized schemes are found by CSA code

Computation system for nuclear reactor core analysis

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

1977-04-01

This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

Analysis of excess reactivity of JOYO MK-III performance test core

International Nuclear Information System (INIS)

Maeda, Shigetaka; Yokoyama, Kenji

2003-10-01

JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

Three-dimensional computer code for the nonlinear dynamic response of an HTGR core

International Nuclear Information System (INIS)

Subudhi, M.; Lasker, L.; Koplik, B.; Curreri, J.; Goradia, H.

1979-01-01

A three-dimensional dynamic code has been developed to determine the nonlinear response of an HTGR core. The HTGR core consists of several thousands of hexagonal core blocks. These are arranged inlayers stacked together. Each layer contains many core blocks surrounded on their outer periphery by reflector blocks. The entire assembly is contained within a prestressed concrete reactor vessel. Gaps exist between adjacent blocks in any horizontal plane. Each core block in a given layer is connected to the blocks directly above and below it via three dowell pins. The present analystical study is directed towards an invesstigation of the nonlinear response of the reactor core blocks in the event of a seismic occurrence. The computer code is developed for a specific mathemtical model which represents a vertical arrangement of layers of blocks. This comprises a block module of core elements which would be obtained by cutting a cylindrical portion consisting of seven fuel blocks per layer. It is anticipated that a number of such modules properly arranged could represent the entire core. Hence, the predicted response of this module would exhibit the response characteristics of the core

Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

International Nuclear Information System (INIS)

Eom, Shin; Oh, Seung-Jong; Diab, Aya

2018-01-01

The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

Energy Technology Data Exchange (ETDEWEB)

Eom, Shin; Oh, Seung-Jong; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering

2018-02-15

The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

A SAS2H/KENO-V Methodology for 3D Full Core depletion analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.; Petrovic, B.

2003-04-01

This paper describes the use of a SAS2H/KENO-V methodology for 3D full core depletion analysis and illustrates its capabilities by applying it to burnup analysis of the IRIS core benchmarks. This new SAS2H/KENO-V sequence combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes. This approach reduces by more than an order of magnitude the time required for getting comparable results using the MOCUP code system. The SAS2H/KENO-V results for the asymmetric IRIS core benchmark are in good agreement with the results of the ALPHA/PHOENIX/ANC code system. (author)

Regional overpower protection system analysis for a DUPIC fuel CANDU core

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok; Park, Jee Won

2003-06-01

The regional overpower protection (ROP) system was assessed a CANDU 6 reactor with the DUPIC fuel, including the validation of the WIMS/RFSP/ROVER-F code system used for the estimation of ROP trip setpoint. The validation calculation has shown that it is valid to use the WIMS/RFSP/ROVER-F code system for ROP system analysis of the CANDU 6 core. For the DUPIC core, the ROP trip setpoint was estimated to be 125.7%, which is almost the same as that of the standard natural uranium core. This study has shown that the DUPIC fuel does not hurt the current ROP trip setpoint designed for the natural uranium CANDU 6 reactor

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

Energy Technology Data Exchange (ETDEWEB)

Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

1983-08-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

International Nuclear Information System (INIS)

Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

1983-01-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Development of core design technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Kim Young In; Kim, Young Il; Kim, Y. G.; Kim, S. J.; Song, H.; Kim, T. K.; Kim, W. S.; Hwang, W.; Lee, B. O.; Park, C. K.; Joo, H. K.; Yoo, J. W.; Kang, H. Y.; Park, W. S

2000-05-01

For the development of KALIMER (150 MWe) core conceptual design, design evolution and optimization for improved economics and safety enhancement was performed in the uranium metallic fueled equilibrium core design which uses U-Zr binary fuel not in excess of 20 percent enrichment. Utilizing results of the uranium ,metallic fueled core design, the breeder equilibrium core design with breeding ratio being over 1.1 was developed. In addition, utilizing LMR's excellent neutron economy, various core concepts for minor actinide burnup, inherent safety, economics and non-proliferation were realized and its optimization studies were performed. A code system for the LMR core conceptual design has been established through the implementation of needed functions into the existing codes and development of codes. To improve the accuracy of the core design, a multi-dimensional nodal transport code SOLTRAN, a three-dimensional transient code analysis code STEP, MATRA-LMR and ASSY-P for T/H analysis are under development. Through the automation of design calculations for efficient core design, an input generator and several interface codes have been developed. (author)

Development of core design technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; In, Kim Young; Kim, Young Il; Kim, Y G; Kim, S J; Song, H; Kim, T K; Kim, W S; Hwang, W; Lee, B O; Park, C K; Joo, H K; Yoo, J W; Kang, H Y; Park, W S

2000-05-01

For the development of KALIMER (150 MWe) core conceptual design, design evolution and optimization for improved economics and safety enhancement was performed in the uranium metallic fueled equilibrium core design which uses U-Zr binary fuel not in excess of 20 percent enrichment. Utilizing results of the uranium ,metallic fueled core design, the breeder equilibrium core design with breeding ratio being over 1.1 was developed. In addition, utilizing LMR's excellent neutron economy, various core concepts for minor actinide burnup, inherent safety, economics and non-proliferation were realized and its optimization studies were performed. A code system for the LMR core conceptual design has been established through the implementation of needed functions into the existing codes and development of codes. To improve the accuracy of the core design, a multi-dimensional nodal transport code SOLTRAN, a three-dimensional transient code analysis code STEP, MATRA-LMR and ASSY-P for T/H analysis are under development. Through the automation of design calculations for efficient core design, an input generator and several interface codes have been developed. (author)

Xenon oscillation in a large PHWR core (Atucha II type): TRISIC code applicability

International Nuclear Information System (INIS)

Solanilla, Roberto

2000-01-01

A three dimensional nuclear reactor simulation code (TRISIC) was developed many years ago to design a PHWR (pressurizer heavy water reactors - Atucha type) based in the 'source-sink model' (heterogeneous theory). The limited processor computational performance available at that time was the constraint of the code when a detailed reactor description was necessary. A modern PC (pentium) code version with a full reactor core representation (461 fuel channels) including diagonal control rod banks and flux-reading detectors with theirs tube guide was used in the present paper for simulation of the Xenon transient when a local asymmetric perturbation was produced in a large core (Atucha II type). The results obtained and the computer time required for the 70 hour's simulation with an adequate time step, established the potential of the code to deal with this kind of transients. The paper shows that the method of TRISIC allows to detect and control azimuthal, radial and axial oscillation. This code is a proper way to elaborate a program of control rods movement from the flux reading detectors to damp the oscillation. TRISIC could also be a accurate tool to supervise the full core flux distribution in real time during the operation of the reactor. (author)

3D thermal-hydraulic analysis on core of PWR nuclear power station

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-09-01

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Directory of Open Access Journals (Sweden)

Dong Hun Lee

2017-01-01

Full Text Available A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

Analysis of loss of coolant accident and emergency core cooling system

International Nuclear Information System (INIS)

Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

1977-01-01

In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

Energy Technology Data Exchange (ETDEWEB)

Chenu, A., E-mail: aurelia.chenu@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Adams, R., E-mail: robert.adams@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Eidgenossische Technische Hochschule, Zurich (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland)

2011-07-01

This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

Analysis of Phenix end-of-life natural convection test with the MARS-LMR code

International Nuclear Information System (INIS)

Jeong, H. Y.; Ha, K. S.; Lee, K. L.; Chang, W. P.; Kim, Y. I.

2012-01-01

The end-of-life test of Phenix reactor performed by the CEA provided an opportunity to have reliable and valuable test data for the validation and verification of a SFR system analysis code. KAERI joined this international program for the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main objectives of this study were to evaluate the capability of existing SFR system analysis code MARS-LMR and to identify any limitation of the code. The analysis was performed in three stages: pre-test analysis, blind posttest analysis, and final post-test analysis. In the pre-test analysis, the design conditions provided by the CEA were used to obtain a prediction of the test. The blind post-test analysis was based on the test conditions measured during the tests but the test results were not provided from the CEA. The final post-test analysis was performed to predict the test results as accurate as possible by improving the previous modeling of the test. Based on the pre-test analysis and blind test analysis, the modeling for heat structures in the hot pool and cold pool, steel structures in the core, heat loss from roof and vessel, and the flow path at core outlet were reinforced in the final analysis. The results of the final post-test analysis could be characterized into three different phases. In the early phase, the MARS-LMR simulated the heat-up process correctly due to the enhanced heat structure modeling. In the mid phase before the opening of SG casing, the code reproduced the decrease of core outlet temperature successfully. Finally, in the later phase the increase of heat removal by the opening of the SG opening was well predicted with the MARS-LMR code. (authors)

Development of severe accident analysis code - Development of a finite element code for lower head failure analysis

Energy Technology Data Exchange (ETDEWEB)

Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-08-01

The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

International Nuclear Information System (INIS)

Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

2017-01-01

Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

Energy Technology Data Exchange (ETDEWEB)

Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

2017-04-15

Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

The APR1400 Core Design by Using APA Code System

International Nuclear Information System (INIS)

Choi, Yu Sun; Koh, Byung Marn

2008-01-01

The nuclear design for APR1400 has been performed to prepare the core model for Automatic Load Follow Operation Simulation. APA (ALPHA/ PHOENIXP/ ANC) code system is a tool for the multi-cycle depletion calculations for APR1400. Its detail versions for ALPHA, PHOENIX-P and ANC are 8.9.3, 8.6.1 and 8.10.5, respectively. The first and equilibrium core depletion calculations for APR1400 have been performed to assure the target cycle length and confirm the safety parameters. The parameters are satisfied within limitation about nuclear design criteria. This APR1400 core models will be based on the design parameters for APR1400 Simulator

FARO base case post-test analysis by COMETA code

Energy Technology Data Exchange (ETDEWEB)

Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

1995-09-01

The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

International Nuclear Information System (INIS)

Mantourov, Guennadi

2001-05-01

This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

International Nuclear Information System (INIS)

Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

1981-02-01

This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

Elastic-plastic stress analysis and ASME code evaluation of a bottomhead penetration in a reactor pressure vessel

International Nuclear Information System (INIS)

Ranganath, S.

1979-01-01

Nuclear pressure vessel components are designed to meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Specifically, the design must satisfy the limits on stress range and fatigue usage prescribed in NB-3200, Section III ASME Code for the various design and operating conditions for the component. The Code requirements assure that the component does not experience gross yielding and that in general, elastic shakedown occurs following cyclic loading. When elastic stress analysis is performed this can be shown by meeting the limits in the Code on Primary and Primary plus Secondary (P+Q) stress intensities. However, when the P+Q limits cannot be met and elastic Shakedown cannot be demonstrated, plastic analysis may be performed to meet the requirements of the Code. This paper describes the elastic-plastic stress analysis of a Boiling Water Reactor Vessel bottom head in-core penetration and illustrates how plastic analysis can be used in ASME Code evaluations to show Code compliance. Details of the thermal analysis, elastic-plastic stress analysis and fatigue evaluation are presented and it is shown that the in-core penetration satisfies the code requirements. 6 refs

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

2014-01-01

Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

Analysis of closed-pool boilup using the TRANSIT-HYDRO code

International Nuclear Information System (INIS)

Graff, D.L.

1983-01-01

The benign termination of the transition phase of a hypothetical LMFBR accident rests on the avoidance of highly energetic recriticalities prior to escape of bottled molten core materials from the active core region. In scenarios where molten fuel is trapped due to axial blockages, the maintenance of subcritical configurations until radial flow paths develop requires stable boil-up of the molten fuel/steel mixture. This paper describes the analysis of an experiment investigating the behavior of closed boiling pools using the two-fluid hydrodynamics module of TRANSIT-HYDRO, a deterministic transition-phase analysis code

A three-dimensional computer code for the nonlinear dynamic response of an HTGR core

International Nuclear Information System (INIS)

Subudhi, M.; Lasker, L.; Koplik, B.; Curreri, J.; Goradia, H.

1979-01-01

A three-dimensional dynamic code has been developed to determine the nonlinear response of an HTGR core. The HTGR core consists of several thousands of hexagonal core blocks. These are arranged in layers stacked together. Each layer contains many core blocks surrounded on their outer periphery by reflector blocks. The entire assembly is contained within a prestressed concrete reactor vessel. Gaps exist between adjacent blocks in any horizontal plane. Each core block in a given layer is connected to the blocks directly above and below it via three dowell pins. The present analytical study is directed towards an investigation of the nonlinear response of the reactor core blocks in the event of a seismic occurrence. The computer code is developed for a specific mathematical model which represents a vertical arrangement of layers of blocks. This comprises a 'block module' of core elements which would be obtained by cutting a cylindrical portion consisting of seven fuel blocks per layer. It is anticipated that a number of such modules properly arranged could represent the entire core. Hence, the predicted response of this module would exhibit the response characteristics of the core. (orig.)

Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

The Analysis of SBWR Critical Power Bundle Using Cobrag Code

Directory of Open Access Journals (Sweden)

Yohannes Sardjono

2013-03-01

Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

Verification of MVP-II and SRAC2006 code to the core physics vera benchmark problem

International Nuclear Information System (INIS)

Jati Susilo

2014-01-01

In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT) of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by. Westinghouse, arranged from 193 unit of 17 x 17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC) and hot zero power (HZP). In this calculation, MVP-II and CITATION module of SRAC2006 computer code has been used with ENDF/B-VII.0. cross section data library. Calculation result showed that differences value of k-eff for the core at controlled and uncontrolled condition between reference with MVP-II (-0,07% and -0,014%) and SRAC2006 (0,92% and 0,99%) are very small or below 1%. Differences value of radial power peaking factor at controlled and uncontrolled of the core between reference value with MVP-II are 0,38% and 1,53%, even though with SRAC2006 are 1,13% and -2,45%. It can be said that the calculation result by both computer code showing suitability with reference value. In order to determinate of criticality of the core, the calculation result using MVP-II code is more conservative compare with SRAC2006 code. (author)

Computation system for nuclear reactor core analysis. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

1977-04-01

This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes

International Nuclear Information System (INIS)

Tramm, J.R.; Siegel, A.R.

2013-01-01

The simulation of whole nuclear cores through the use of Monte Carlo codes requires an impracticably long time-to-solution. We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm - the calculation of macroscopic cross sections - in an effort to expose bottlenecks within multi-core, shared memory architectures. (authors)

Selection and benchmarking of computer codes for research reactor core conversions

International Nuclear Information System (INIS)

Yilmaz, E.; Jones, B.G.

1983-01-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of Illinois. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general

Verification of computer code for calculation of coolant radiolysis in the VVER reactor core with regard for boiling in its upper part

Energy Technology Data Exchange (ETDEWEB)

Arkhipov, O.P.; Kabakchi, S.A. [OKB Gidropress, Podolsk, Moscow (Russian Federation)

2010-07-01

Code Bora for WWER coolant radiolysis calculation considering single jets boiling in the reactor core top part is developed on the basis of computer codes MOPABA-H2 (radiolysis of aqueous solutions) and SteamRad (radiolysis of vapor). Physico-chemical processes taking place in boiling core coolant are complex and diversified. Still, for the solution of certain problems their simulation can be simplified. The approach of reasonable simplification was used for development of code Bora: mathematical model assumed is purposed for simulation of phenomena only in the area of interest; the number of simulated chemical reactions and particles shall be reasonably minimum; complexity of interphase mass transfer calculation procedure shall be adequate to actually available accuracy of modeling. The analysis of new experimental initial yields of water radiolysis products data and kinetic parameters of elementary chemical reactions with their participation has been carried out. Some changes have been introduced in the mechanism of liquid water and aqueous solutions of ammonia radiolysis have been significantly revised on the basis of this analysis. Examples of the calculations provided for code Bora verification are presented. Despite of very simple simulation of interphase mass transfer, Bora allows to obtain average chemical composition of two-phase coolant at BWR core outlet with the accuracy sufficient for engineering calculations. The report also presents the results of two-phase coolant chemical composition test calculation for reactor core top part coolant boiling in pressurized water reactor. (author)

Introduction of SCIENCE code package

International Nuclear Information System (INIS)

Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

2012-01-01

The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

User's manual for seismic analysis code 'SONATINA-2V'

Energy Technology Data Exchange (ETDEWEB)

Hanawa, Satoshi; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2001-08-01

The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

Thermal-hydraulic analysis of PWR cores in transient condition

International Nuclear Information System (INIS)

Silva Galetti, M.R. da.

1984-01-01

A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

Evaluation of a thermal SCWR core with sub-channel analysis

International Nuclear Information System (INIS)

Liu Xiaojing; Cheng Xu

2008-01-01

A previous study shows that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behaviour than the existing one-row fuel assemblies. With this new developed two-row fuel assembly, a thermal SCWR core design is proposed Assessment of this design is carried out in this paper. The performance of this new core design is investigated with 3-D coupled thermal-hydraulic/neutronic calculations. During the coupling procedure, the thermal-hydraulic behaviour is analyzed using a single-channel code and the neutron-physical performance is computed with a 3-D reactor physical code. This paper presents the main results achieved so far related to the distribution of some neutronic and thermal-hydraulic parameters. Since the power distribution in some fuel assemblies is extremely uneven, sub-channel analysis is applied to the hottest and most non-uniform assembly in the core. The sub-channel analysis is performed with the power and thermal hydraulic parameters from the coupling results. It provides the hot channel factor and the maximal cladding surface temperature more precisely. The power and mass flux distribution in these assemblies are illustrated in detail for the demonstration purpose. The difference of the results evaluated with two different methods, i.e. sub-channel analysis and single-channel analysis, shows the importance of applying sub-channel analysis. A sensitivity analysis of some important parameters is also carried out. (author)

Analysis of subchannel effects and their treatment in average channel PWR core models

International Nuclear Information System (INIS)

Cuervo, D.; Ahnert, C.; Aragones, J.M.

2004-01-01

Neutronic thermal-hydraulic coupling is meanly made at this moment using whole plant thermal-hydraulic codes with one channel per assembly or quarter of assembly in more detailed cases. To extract safety limits variables a new calculation has to be performed using thermal-hydraulic subchannel codes in an embedded or off-line manner what implies an increase of calculation time. Another problem of this separated analysis of whole core and not channel is that the whole core calculation is not resolving the real problem due to the modification of the variables values by the homogenization process that is carried out to perform the whole core analysis. This process is making that some magnitudes are over or under-predicted causing that the problem that is being solved is not the original one. The purpose of the work that is being developed is to investigate the effects of the averaging process in the results obtained by the whole core analysis and to develop some corrections that may be included in this analysis to obtain results closer to the ones obtained by a detailed subchannel analysis. This paper shows the results obtained for a sample case and the conclusions for future work. (author)

Validation of a Subchannel Analysis Code MATRA Version 1.0

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Seo, Kyung Won; Kwon, Hyouk

2008-10-15

A subchannel analysis code MATRA has been developed for the thermal hydraulic analysis of SMART core. The governing equations and important models were established, and validation calculations have been performed for subchannel flow and enthalpy distributions in rod bundles under steady-state conditions. The governing equations of the MATRA were on the basis of integral balance equation of the two-phase mixture. The effects of non-homogeneous and non-equilibrium states were considered by employing the subcooled boiling model and the phasic slip model. Solution scheme and main structure of the MATRA code, as well as the difference of MATRA and COBRA-IV-I codes, were summarized. Eight different test data sets were employed for the validation of the MATRA code. The collected data consisted of single-phase subchannel flow and temperature distribution data, single-phase inlet flow maldistribution data, single-phase partial flow blockage data, and two-phase subchannel flow and enthalpy distribution data. The prediction accuracy as well as the limitation of the MATRA code was evaluated from this analysis.

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

Preliminary Uncertainty Analysis for SMART Digital Core Protection and Monitoring System

International Nuclear Information System (INIS)

Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun

2012-01-01

The Korea Atomic Energy Research Institute (KAERI) developed on-line digital core protection and monitoring systems, called SCOPS and SCOMS as a part of SMART plant protection and monitoring system. SCOPS simplified the protection system by directly connecting the four RSPT signals to each core protection channel and eliminated the control element assembly calculator (CEAC) hardware. SCOMS adopted DPCM3D method in synthesizing core power distribution instead of Fourier expansion method being used in conventional PWRs. The DPCM3D method produces a synthetic 3-D power distribution by coupling a neutronics code and measured in-core detector signals. The overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system was developed. In this paper, preliminary overall uncertainty factors for SCOPS/SCOMS of SMART initial core were evaluated by applying newly developed uncertainty analysis method

RAP-2A Computer code for transients analysis in fast reactors

International Nuclear Information System (INIS)

Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.

1975-10-01

The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

Melt spreading code assessment, modifications, and application to the EPR core catcher design

International Nuclear Information System (INIS)

Farmer, M.T.

2009-01-01

of: (1) comparison to an analytical solution for the dam break problem, (2) water spreading tests in a 1/10 linear scale model of the Mark I containment by Theofanous et al., and (3) steel spreading tests by Suzuki et al. that were also conducted in a geometry similar to the Mark I. The objective of this work was to utilize the MELTSPREAD code to check the assumption of uniform melt spreading in the EPR core catcher design. As a starting point for the project, the code was validated against the worldwide melt spreading database that emerged after the code was originally written in the very early 1990's. As part of this exercise, the code was extensively modified and upgraded to incorporate findings from these various analytical and experiment programs. In terms of expanding the ability of the code to analyze various melt simulant experiments, the options to input user-specified melt and/or substrate material properties was added. The ability to perform invisicid and/or adiabatic spreading analysis was also added so that comparisons with analytical solutions and isothermal spreading tests could be carried out. In terms of refining the capability to carry out reactor material melt spreading analyses, the code was upgraded with a new melt viscosity model; the capability was added to treat situations in which solid fraction buildup between the liquidus-solidus is non-linear; and finally, the ability to treat an interfacial heat transfer resistance between the melt and substrate was incorporated. This last set of changes substantially improved the predictive capability of the code in terms of addressing reactor material melt spreading tests. Aside from improvements and upgrades, a method was developed to fit the model to the various melt spreading tests in a manner that allowed uncertainties in the model predictions to be statistically characterized. With these results, a sensitivity study was performed to investigate the assumption of uniform spreading in the EPR core

Automated and Assistive Tools for Accelerated Code migration of Scientific Computing on to Heterogeneous MultiCore Systems

Science.gov (United States)

2017-04-13

AFRL-AFOSR-UK-TR-2017-0029 Automated and Assistive Tools for Accelerated Code migration of Scientific Computing on to Heterogeneous MultiCore Systems ...2012, “ Automated and Assistive Tools for Accelerated Code migration of Scientific Computing on to Heterogeneous MultiCore Systems .” 2. The objective...2012 - 01/25/2015 4. TITLE AND SUBTITLE Automated and Assistive Tools for Accelerated Code migration of Scientific Computing on to Heterogeneous

Exposure calculation code module for reactor core analysis: BURNER

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

Exposure calculation code module for reactor core analysis: BURNER

International Nuclear Information System (INIS)

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

Development of the computer code system for the analyses of PWR core

International Nuclear Information System (INIS)

Tsujimoto, Iwao; Naito, Yoshitaka.

1992-11-01

This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores

International Nuclear Information System (INIS)

Raymond, P.; Allaire, G.; Boudsocq, G.

1995-01-01

FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties

Computer code for simulating pressurized water reactor core

International Nuclear Information System (INIS)

Serrano, A.M.B.

1978-01-01

A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

International Nuclear Information System (INIS)

Rocek, J.; Belac, J.; Miasnikov, A.

2003-01-01

The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

THYDE-P2 code: RCS (reactor-coolant system) analysis code

International Nuclear Information System (INIS)

Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

1986-12-01

THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)

The effects of core zoning on optimization of design analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

Preliminary Coupling of MATRA Code for Multi-physics Analysis

International Nuclear Information System (INIS)

Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

2014-01-01

The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

DABIE: a data banking system of integral experiments for reactor core characteristics computer codes

International Nuclear Information System (INIS)

Matsumoto, Kiyoshi; Naito, Yoshitaka; Ohkubo, Shuji; Aoyanagi, Hideo.

1987-05-01

A data banking system of integral experiments for reactor core characteristics computer codes, DABIE, has been developed to lighten the burden on searching so many documents to obtain experiment data required for verification of reactor core characteristics computer code. This data banking system, DABIE, has capabilities of systematic classification, registration and easy retrieval of experiment data. DABIE consists of data bank and supporting programs. Supporting programs are data registration program, data reference program and maintenance program. The system is designed so that user can easily register information of experiment systems including figures as well as geometry data and measured data or obtain those data through TSS terminal interactively. This manual describes the system structure, how-to-use and sample uses of this code system. (author)

Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

Energy Technology Data Exchange (ETDEWEB)

Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

1995-09-01

The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

Core Seismic Tests for a Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, Gyeong Hoi; Lee, J. H

2007-01-15

This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

International Nuclear Information System (INIS)

Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

2013-01-01

The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

Interface of RETRAN/MASTER Code System for APR1400

International Nuclear Information System (INIS)

Ku, Keuk Jong; Kang, Sang Hee; Kim, Han Gon

2008-01-01

MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is a nuclear analysis and design code which can simulate the pressurized water reactor core or boiling water reactor core in 3-dimensional geometry. RETRAN is a best-estimate code for transient analysis of Non-LOCA. RETRAN code generates neutron number density in core using point kinetics model which includes feedback reactivities and converts the neutron number density into reactor power. It is conventional that RETRAN code for power generation is roughly to extrapolate feedback reactivities which are provided by MASTER code only one time before transient analysis. The purpose of this paper is to interface RETRAN code with MASTER code by real-time processing and to supply adequate feedback reactivities to RETRAN code. So, we develop interface code called MATRAN for real-time feedback reactivity processing. And for the application of MATRAN code, we compare the results of real-time MATRAN code with those of conventional RETRAN/MASTER code

Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.

2006-01-01

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Seo, K. W

2006-01-15

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

Optimization of multi-phase compressible lattice Boltzmann codes on massively parallel multi-core systems

NARCIS (Netherlands)

Biferale, L.; Mantovani, F.; Pivanti, M.; Pozzati, F.; Sbragaglia, M.; Schifano, S.F.; Toschi, F.; Tripiccione, R.

2011-01-01

We develop a Lattice Boltzmann code for computational fluid-dynamics and optimize it for massively parallel systems based on multi-core processors. Our code describes 2D multi-phase compressible flows. We analyze the performance bottlenecks that we find as we gradually expose a larger fraction of

Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

International Nuclear Information System (INIS)

Xhonneux, Andre; Allelein, Hans-Josef

2014-01-01

The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

Verification of the DEFENS Code through the CANDU Problems with Rectangular Geometry

Energy Technology Data Exchange (ETDEWEB)

Ryu, Eun Hyun; Song, Yong Mann [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Because a finite element method (FEM) based code can explicitly describe the core geometry, it has an advantage in a core analysis such as the CANDU core. For the reactor physics calculation in the CANDU core, the RFSP-IST code is used for the core analysis, and the RFSP-IST code is based on the finite difference method (FDM). Thus, the convergence with the mesh size and the geometry shape is not consistent. In this research, the convergence with the mesh size of the RFSP code is investigated, a method comparison between the FEM and FDM is done for the usefulness of the FEM based code with the same rectangular geometry. The target problems are the imaginary core and initial core with the uniform parameter, which is produced by the WIMS-IST code based on the parameters of Wolsong unit 1. The reference solution is generated by running the multi-group calculation of the McCARD code. In this research, the convergence of the RFSP code is investigated and the DEFENS code is compared with the RFSP code for the imaginary and initial cores. The accuracy of the DEFENS code and the disadvantage of the RFSP code are verified.

Subchannel analysis of a boiloff experiment by a system thermalhydraulic code

International Nuclear Information System (INIS)

Bousbia-Salah, A.; D'Auria, F.

2001-01-01

This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the Neptun low pressure test No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

GENOVA: a generalized perturbation theory program for various applications to CANDU core physics analysis (I)-theory and application

International Nuclear Information System (INIS)

Kim, Do Heon; Choi, Hang Bok

2001-01-01

A generalized perturbation theory (GPT) program, GENOVA, has been developed for the purpose of various applications to Canadian deuterium uranium (CANDU) reactor physics analyses. GENOVA was written under the framework of CANDU physics design and analysis code, RFSP. A sensitivity method based on the GPT was implemented in GENOVA to estimate various sensitivity coefficients related to the movement of zone controller units (ZCUs) existing in the CANDU reactor. The numerical algorithm for the sensitivity method was verified by a simple 2 x 2 node problem. The capability of predicting ZCU levels upon a refueling perturbation was validated for a CANDU-6 reactor problem. The applicability of GENOVA to the CANDU-6 core physics analysis has been demonstrated with the optimum refueling simulation and the uncertainty analysis problems. For the optimum refueling simulation, an optimum channel selection strategy has been proposed, using the ZCU level predicted by GENOVA. The refueling simulation of a CANDU-6 natural uranium core has shown that the ZCU levels are successfully controlled within the operating range while the channel and bundle powers are satisfying the license limits. An uncertainty analysis has been performed for the fuel composition heterogeneity of a CANDU DUPIC core, using the sensitivity coefficients generated by GENOVA. The results have shown that the uncertainty of the core performance parameter can be reduced appreciably when the contents of the major fissile isotopes are tightly controlled. GENOVA code has been successfully explored to supplement the weak points of the current design and analysis code, such as the incapacity of performing an optimum refueling simulation and uncertainty analysis. The sample calculations have shown that GENOVA has strong potential to be used for CANDU core analysis combined with the current design and analysis code, RFSP, especially for the development of advanced CANDU fuels

Analysis of a basic core performance for FBR core nuclear design. 3

International Nuclear Information System (INIS)

Kaneko, Kunio

1999-03-01

The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

European ERANOS formulaire for fast reactor core analysis

International Nuclear Information System (INIS)

Rimpault, Gerald

2003-01-01

ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

Selection and benchmarking of computer codes for research reactor core conversions

Energy Technology Data Exchange (ETDEWEB)

Yilmaz, Emin [School of Aerospace, Mechanical and Nuclear Engineering, University of Oklahoma, Norman, OK (United States); Jones, Barclay G [Nuclear Engineering Program, University of IL at Urbana-Champaign, Urbana, IL (United States)

1983-09-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC{sup 2}, COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k{sub eff} is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

Selection and benchmarking of computer codes for research reactor core conversions

International Nuclear Information System (INIS)

Yilmaz, Emin; Jones, Barclay G.

1983-01-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k eff is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

ANDREA: Advanced nodal diffusion code for reactor analysis

International Nuclear Information System (INIS)

Belac, J.; Josek, R.; Klecka, L.; Stary, V.; Vocka, R.

2005-01-01

A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Ustun, G.; Durmayaz, A.

2002-01-01

Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

Two dimensional dynamic analysis of sandwich plates with gradient foam cores

Energy Technology Data Exchange (ETDEWEB)

Mu, Lin; Xiao, Deng Bao; Zhao, Guiping [State Key Laboratory for Mechanical structure Strength and Vibration, School of AerospaceXi' an Jiaotong University, Xi' an (China); Cho, Chong Du [Dept. of Mechanical Engineering, Inha University, Inchon (Korea, Republic of)

2016-09-15

Present investigation is concerned about dynamic response of composite sandwich plates with the functionally gradient foam cores under time-dependent impulse. The analysis is based on a model of the gradient sandwich plate, in which the face sheets and the core adopt the Kirchhoff theory and a [2, 1]-order theory, respectively. The material properties of the gradient foam core vary continuously along the thickness direction. The gradient plate model is validated with the finite element code ABAQUS®. And the results show that the proposed model can predict well the free vibration of composite sandwich plates with gradient foam cores. The influences of gradient foam cores on the natural frequency, deflection and energy absorbing of the sandwich plates are also investigated.

BEAVRS full core burnup calculation in hot full power condition by RMC code

International Nuclear Information System (INIS)

Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

2017-01-01

Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

REFLA-1D/MODE3: a computer code for reflood thermo-hydrodynamic analysis during PWR-LOCA

International Nuclear Information System (INIS)

Murao, Yoshio; Okubo, Tsutomu; Sugimoto, Jun; Iguchi, Tadashi; Sudoh, Takashi.

1985-02-01

This manual describes the REFLA-1D/MODE3 reflood system analysis code. This code can solve the core thermo-hydrodynamics under forced flooding conditions and gravity feed conditions in a system similar to FLECHT-SET Phase A. This manual describes the REFLA-1D/MODE3 models and provides application information required to utilize the code. (author)

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

1978-01-01

The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

ARTEMIS: The core simulator of AREVA NP's next generation coupled neutronics/thermal-hydraulics code system ARCADIAR

International Nuclear Information System (INIS)

Hobson, Greg; Merk, Stephan; Bolloni, Hans-Wilhelm; Breith, Karl-Albert; Curca-Tivig, Florin; Van Geemert, Rene; Heinecke, Jochen; Hartmann, Bettina; Porsch, Dieter; Tiles, Viatcheslav; Dall'Osso, Aldo; Pothet, Baptiste

2008-01-01

AREVA NP has developed a next-generation coupled neutronics/thermal-hydraulics code system, ARCADIA R , to fulfil customer's current demands and even anticipate their future demands in terms of accuracy and performance. The new code system will be implemented world-wide and will replace several code systems currently used in various global regions. An extensive phase of verification and validation of the new code system is currently in progress. One of the principal components of this new system is the core simulator, ARTEMIS. Besides the stand-alone tests on the individual computational modules, integrated tests on the overall code are being performed in order to check for non-regression as well as for verification of the code. Several benchmark problems have been successfully calculated. Full-core depletion cycles of different plant types from AREVA's French, American and German regions (e.g. N4 and KONVOI types) have been performed with ARTEMIS (using APOLLO2-A cross sections) and compared directly with current production codes, e.g. with SCIENCE and CASCADE-3D, and additionally with measurements. (authors)

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR

International Nuclear Information System (INIS)

Wang, S.J.; Chien, C.S.

1996-01-01

The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required

Methodology for reactor core physics analysis - part 2; Metodologia de analise fisica do nucleo - etapa 2

Energy Technology Data Exchange (ETDEWEB)

Ponzoni Filho, P; Fernandes, V B; Lima Bezerra, J de; Santos, T I.C.

1992-12-01

The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs.

Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

Directory of Open Access Journals (Sweden)

Aldawahra Saadou

2015-06-01

Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

International Nuclear Information System (INIS)

Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

2005-01-01

In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET

International Nuclear Information System (INIS)

Hainoun, A.; Alhabet, F.

2004-02-01

Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed

Computer code and users' guide for the preliminary analysis of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239b

Energy Technology Data Exchange (ETDEWEB)

Nichols, R.A.; Smith, W.W.

1976-06-30

The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The second volume describes the computer code and users' guide for the preliminary analysis of the system.

Steady state thermal hydraulic analysis of a boiling water reactor core, for various power distributions, using computer code THABNA

International Nuclear Information System (INIS)

Venkat Raj, V.; Saha, D.

1976-01-01

The core of a boiling water reactor may see different power distributions during its operational life. How some of the typical power distributions affect some of the thermal hydraulic parameters such as pressure drop minimum critical heat flux ratio, void distribution etc. has been studied using computer code THABNA. The effect of an increase in the leakage flow has also been analysed. (author)

ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

International Nuclear Information System (INIS)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

2007-01-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

Melt spreading code assessment, modifications, and initial application to the EPR core catcher design

International Nuclear Information System (INIS)

Farmer, M.T.; Basu, S.

2009-01-01

The Evolutionary Power Reactor (EPR) is a 1,600-MWe Pressurized Water Reactor (PWR) that is undergoing a design certification review by the U.S. Nuclear Regulatory Commission (NRC). The EPR severe accident design philosophy is predicated upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external flooding. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: 1) an external core melt retention system to temporarily hold core melt released from the vessel; 2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; 3) a melt plug that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, 4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and non-uniform spreading. The NRC is using MELTSPREAD to evaluate melt spreading in the EPR design. The development of MELTSPREAD ceased in the early 1990's, and so the code was first assessed against the more contemporary spreading database and code modifications, as warranted, were carried out before performing confirmatory plant calculations. This paper provides principle findings from the MELTSPREAD assessment activities and resulting code modifications, and also summarizes the results of initial scoping calculations for the EPR plant design and preliminary plant analyses, along with the plan for performing the final set of plant calculations including sensitivity studies

Development status of Severe Accident Analysis Code SAMPSON

International Nuclear Information System (INIS)

Iwashita, Tsuyoshi; Ujita, Hiroshi

2000-01-01

The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

International Nuclear Information System (INIS)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

2017-01-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

Energy Technology Data Exchange (ETDEWEB)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

Installation of JMTR core management system

International Nuclear Information System (INIS)

Imaizumi, Tomomi; Ide, Hiroshi; Naka, Michihiro; Komukai, Bunsaku; Nagao, Yoshiharu

2013-01-01

In order to carry out the core management after the reoperation of JMTR quickly and accurately, the authors took up the Standard Reactor Analysis Code (SRAC) system and core management support programs that are operating in a general-purpose large computer and transferred them to PC (OS: Linux), and newly established a JMTR core management system. As for the core analysis, this measure enabled an increase in the processing speed from the check of core arrangement to the result display of nuclear restriction values to about 60 times, compared with the conventional method. It was confirmed that the differences of calculation results originated from the difference of internal display of computers, associated with the transfer of each analysis code from GS21-400 system to PC-Linux, were within practically allowable level. In the future, this system will be applied to the core analysis of JMTR, as well as to the preparation of operation plans. (A.O.)

Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

Directory of Open Access Journals (Sweden)

Reza Akbari

2017-08-01

Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

TASS code topical report. V.1 TASS code technical manual

International Nuclear Information System (INIS)

Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

1997-02-01

TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

Status of the development of a fully integrated code system for the simulation of high temperature reactor cores

Energy Technology Data Exchange (ETDEWEB)

Kasselmann, Stefan, E-mail: s.kasselmann@fz-juelich.de [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Druska, Claudia [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Herber, Stefan [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Jühe, Stephan [Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany); Keller, Florian; Lambertz, Daniela; Li, Jingjing; Scholthaus, Sarah; Shi, Dunfu [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Xhonneux, Andre; Allelein, Hans-Josef [Institute of Energy and Climate Research, Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Jülich GmbH, 52425 Jülich (Germany); Lehrstuhl für Reaktorsicherheit und -technik, RWTH Aachen, 52062 Aachen (Germany)

2014-05-01

The HTR code package (HCP) is a new code system, which couples a variety of stand-alone codes for the simulation of different aspects of HTR. HCP will allow the steady-state and transient operating conditions of a 3D reactor core to be simulated including new features such as spatially resolved fission product release calculations or production and transport of graphite dust. For this code the latest programming techniques and standards are applied. As a first step an object-oriented data model was developed which features a high level of readability because it is based on problem-specific data types like Nuclide, Reaction, ReactionHandler, CrossSectionSet, etc. Those classes help to encapsulate and therefore hide specific implementations, which are not relevant with respect to physics. HCP will make use of one consistent data library for which an automatic generation tool was developed. The new data library consists of decay information, cross sections, fission yields, scattering matrices etc. for all available nuclides (e.g. ENDF/B-VII.1). The data can be stored in different formats such as binary, ASCII or XML. The new burn up code TNT (Topological Nuclide Transmutation) applies graph theory to represent nuclide chains and to minimize the calculation effort when solving the burn up equations. New features are the use of energy-dependent fission yields or the calculation of thermal power for decay, fission and capture reactions. With STACY (source term analysis code system) the fission product release for steady state as well as accident scenarios can be simulated for each fuel batch. For a full-core release calculation several thousand fuel elements are tracked while passing through the core. This models the stochastic behavior of a pebble bed in a realistic manner. In this paper we report on the current status of the HCP and present first results, which prove the applicability of the selected approach.

Benchmarking Data Analysis and Machine Learning Applications on the Intel KNL Many-Core Processor

OpenAIRE

Byun, Chansup; Kepner, Jeremy; Arcand, William; Bestor, David; Bergeron, Bill; Gadepally, Vijay; Houle, Michael; Hubbell, Matthew; Jones, Michael; Klein, Anna; Michaleas, Peter; Milechin, Lauren; Mullen, Julie; Prout, Andrew; Rosa, Antonio

2017-01-01

Knights Landing (KNL) is the code name for the second-generation Intel Xeon Phi product family. KNL has generated significant interest in the data analysis and machine learning communities because its new many-core architecture targets both of these workloads. The KNL many-core vector processor design enables it to exploit much higher levels of parallelism. At the Lincoln Laboratory Supercomputing Center (LLSC), the majority of users are running data analysis applications such as MATLAB and O...

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

Advanced thermionic reactor systems design code

International Nuclear Information System (INIS)

Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

1991-01-01

An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

International Nuclear Information System (INIS)

Araya, F.

1995-01-01

After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

Intercomparison of liquid metal fast reactor seismic analysis codes. V. 2: Verification and improvement of reactor core seismic analysis codes using core mock-up experiments. Proceedings of a research co-ordination meeting held in Vienna, 26-28 September 1994

International Nuclear Information System (INIS)

1995-10-01

This report (Volume II) contains the papers summarizing the verification of and improvement to the codes on the basis of the French and Japanese data. Volume I: ''Validation of the Seismic Analysis Codes Using the Reactor Code Experiments'' (IAEA-TECDOC-798) included the Italian PEC reactor data. Refs, figs and tabs

Intercomparison of liquid metal fast reactor seismic analysis codes. V. 2: Verification and improvement of reactor core seismic analysis codes using core mock-up experiments. Proceedings of a research co-ordination meeting held in Vienna, 26-28 September 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-10-01

This report (Volume II) contains the papers summarizing the verification of and improvement to the codes on the basis of the French and Japanese data. Volume I: ``Validation of the Seismic Analysis Codes Using the Reactor Code Experiments`` (IAEA-TECDOC-798) included the Italian PEC reactor data. Refs, figs and tabs.

PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

1996-09-01

The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)

Improvement of calculation method for temperature coefficient of HTTR by neutronics calculation code based on diffusion theory. Analysis for temperature coefficient by SRAC code system

International Nuclear Information System (INIS)

Goto, Minoru; Takamatsu, Kuniyoshi

2007-03-01

The HTTR temperature coefficients required for the core dynamics calculations had been calculated from the HTTR core calculation results by the diffusion code with which the corrections had been performed using the core calculation results by the Monte-Carlo code MVP. This calculation method for the temperature coefficients was considered to have some issues to be improved. Then, the calculation method was improved to obtain the temperature coefficients in which the corrections by the Monte-Carlo code were not required. Specifically, from the point of view of neutron spectrum calculated by lattice calculations, the lattice model was revised which had been used for the calculations of the temperature coefficients. The HTTR core calculations were performed by the diffusion code with the group constants which were generated by the lattice calculations with the improved lattice model. The core calculations and the lattice calculations were performed by the SRAC code system. The HTTR core dynamics calculation was performed with the temperature coefficient obtained from the core calculation results. In consequence, the core dynamics calculation result showed good agreement with the experimental data and the valid temperature coefficient could be calculated only by the diffusion code without the corrections by Monte-Carlo code. (author)

Feasibility study of applying a multi-channel analysis model to on-line core monitoring system

International Nuclear Information System (INIS)

In, W. K.; Yoo, Y. J.; Hwang, D. H.; Jun, T. H.

1998-01-01

A feasibility study was performed to evaluate the effect of implementing a multi-channel analysis model in on-line core monitoring system. A simplified thermal-hydraulic model has been used in the on-line core monitoring system of digital PWR. The design procedure, core thermal margin and computation time were investigated in case of replacing the simplified model with the multi-channel analysis model. For the given ranges of limiting conditions for operation in Yonggwang Unit 3 Cycle 1, the minimum DNBR of the simplified thermal-hydraulic code CETOP-D was compared to that of the multi-channel analysis code MATRA. A CETOP-D tuning is additionally required to ensure the accurate and conservative DNBR calculation but the MATRA tuning is not necessary. MATRA appeared to increase the DNBR overpower margin from 2.5% to 6% over the CETOP-D margin. MATRA took approximately 1 second to compute DNBR on the HP9000 workstation system, which is longer than the DNBR computation time of CETOP-D. It is, however, fast enough to perform the on-line monitoring of DNBR. It can be therefore concluded that the application of the multi-channel analysis model MATRA in the on-line core monitoring system is feasible

Benchmark Simulation for the Development of the Regulatory Audit Subchannel Analysis Code

Energy Technology Data Exchange (ETDEWEB)

Lee, G. H.; Song, C.; Woo, S. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-05-15

For the safe and reliable operation of a reactor, it is important to predict accurately the flow and temperature distributions in the thermal-hydraulic design of a reactor core. A subchannel approach can give the reasonable flow and temperature distributions with the short computing time. Korea Institute of Nuclear Safety (KINS) is presently reviewing new subchannel code, THALES, which will substitute for both THINC-IV and TORC code. To assess the prediction performance of THALES, KINS is developing the subchannel analysis code for the independent audit calculation. The code is based on workstation version of COBRA-IV-I. The main objective of the present study is to assess the performance of COBRA-IV-I code by comparing the simulation results with experimental ones for the sample problems

Analysis Of The Effect Of Fuel Enrichment Error On Neutronic Properties Of The RSG-GAS Core

International Nuclear Information System (INIS)

Saragih, Tukiran; Pinem, Surian

2002-01-01

The analysis of the fuel enrichment error effect on neutronic properties has been carried out. The fuel enrichment could be improperly done because of wrong fabrication. Therefore it is necessary to analyze the fuel enrichment error effect to determine how many percents the fuel enrichment maximum can be accepted in the core. The analysis was done by simulation method The RSG-GAS core was simulated with 5 standard fuels and 1 control element having wrong enrichment when inserted into the core. Fuel enrichment error was then simulated from 20%, 25% and 30% and the simulation was done using WIMSD/4 and Batan-2DIFF codes. The cross section of core material of the RSG-GAS was generated by WIMSD/4 code in 1-D, X-Y geometry and 10 energy neutron group. Two dimensions, diffusion calculation based on finite element method was done by using Batan-2DIFF code. Five fuel elements and one control element changed the enrichment was finally arranged as a new core of the RSG-Gas reactor. The neutronic properties can be seen from eigenvalues (k eff ) as well as from the kinetic properties based on moderator void reactivity coefficient. The calculated results showed that the error are still acceptable by k eff 1,097 even until 25% fuel enrichment but not more than 25,5%

Analysis of impurity effect on Silicide fuels of the RSG-GAS core

International Nuclear Information System (INIS)

Tukiran-Surbakti

2003-01-01

Simulation of impurity effect on silicide fuel of the RSG-GAS core has been done. The aim of this research is to know impurity effect of the U-234 and U-236 isotopes in the silicide fuels on the core criticality. The silicide fuels of 250 g U loading and 19.75 of enrichment is used in this simulation. Cross section constant of fuels and non-structure material of core are generated by WIMSD/4 computer code, meanwhile impurity concentration was arranged from 0.01% to 2%. From the result of analysis can be concluded that the isotopes impurity in the fuels could make trouble in the core and the core can not be operated at critical after a half of its cycle length (350 MW D)

Multi codes and multi-scale analysis for void fraction prediction in hot channel for VVER-1000/V392

International Nuclear Information System (INIS)

Hoang Minh Giang; Hoang Tan Hung; Nguyen Huu Tiep

2015-01-01

Recently, an approach of multi codes and multi-scale analysis is widely applied to study core thermal hydraulic behavior such as void fraction prediction. Better results are achieved by using multi codes or coupling codes such as PARCS and RELAP5. The advantage of multi-scale analysis is zooming of the interested part in the simulated domain for detail investigation. Therefore, in this study, the multi codes between MCNP5, RELAP5, CTF and also the multi-scale analysis based RELAP5 and CTF are applied to investigate void fraction in hot channel of VVER-1000/V392 reactor. Since VVER-1000/V392 reactor is a typical advanced reactor that can be considered as the base to develop later VVER-1200 reactor, then understanding core behavior in transient conditions is necessary in order to investigate VVER technology. It is shown that the item of near wall boiling, Γ w in RELAP5 proposed by Lahey mechanistic method may not give enough accuracy of void fraction prediction as smaller scale code as CTF. (author)

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

International Nuclear Information System (INIS)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

Energy Technology Data Exchange (ETDEWEB)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

Calculation analysis of TRIGA MARK II reactor core composed of two types of fuel elements

International Nuclear Information System (INIS)

Ravnik, M.

1988-11-01

The most important properties of mixed cores are treated for TRIGA MARK II reactor, composed of standard (20% enriched, 8.5w% U content) and FLIP (70% enriched, 8.5w% U content) fuel elements. Large difference in enrichment and presence of burnable poison in FLIP fuel have strong influence on the main core characteristics, such as: fuel temperature coefficient, power defect, Xe and Sm worth, power and flux distributions, etc. They are significantly different for both types of fuel. Optimal loading of mixed cores therefore strongly depends on the loading pattern of both types of fuel elements. Results of systematic calculational analysis of mixed cores are presented. Calculations on the level of fuel element are performed with WIMSD-4 computer code with extended cross-section library. Core calculations are performed with TRIGAP two-group 1-D diffusion code. Results are compared to measurements and physical explanation is provided. Special concern is devoted to realistic mixed cores, for which optimal in-core fuel management is derived. Refs, figs and tabs

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

Science.gov (United States)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

Code for the core simulation in pressurized water reactors

International Nuclear Information System (INIS)

Serrano, M.A.B.

1978-08-01

A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

SIMMER-II code and its applications

International Nuclear Information System (INIS)

Smith, L.L.

1979-01-01

The significant features of SIMMER-II, a disrupted-core analysis code, are described. The code has the capabalities to begin space-time neutronics calculations from nonstationary reactor states, to track the intermixing of fuel of different enrichments, and to model the complicated heat- and mass-transfer processes that occur in the transition phase. Example calculations are presented for analyses of whole-core accidents and for analyses of experiments supporting the code models

Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

International Nuclear Information System (INIS)

Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

1984-01-01

Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

2D and 3D core-collapse supernovae simulation results obtained with the CHIMERA code

Energy Technology Data Exchange (ETDEWEB)

Bruenn, S W; Marronetti, P; Dirk, C J [Physics Department, Florida Atlantic University, 777 W. Glades Road, Boca Raton, FL 33431-0991 (United States); Mezzacappa, A; Hix, W R [Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Blondin, J M [Department of Physics, North Carolina State University, Raleigh, NC 27695-8202 (United States); Messer, O E B [Center for Computational Sciences, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6354 (United States); Yoshida, S, E-mail: bruenn@fau.ed [Max-Planck-Institut fur Gravitationsphysik, Albert Einstein Institut, Golm (Germany)

2009-07-01

Much progress in realistic modeling of core-collapse supernovae has occurred recently through the availability of multi-teraflop machines and the increasing sophistication of supernova codes. These improvements are enabling simulations with enough realism that the explosion mechanism, long a mystery, may soon be delineated. We briefly describe the CHIMERA code, a supernova code we have developed to simulate core-collapse supernovae in 1, 2, and 3 spatial dimensions. We then describe the results of an ongoing suite of 2D simulations initiated from a 12, 15, 20, and 25 M{sub o-dot} progenitor. These have all exhibited explosions and are currently in the expanding phase with the shock at between 5,000 and 20,000 km. We also briefly describe an ongoing simulation in 3 spatial dimensions initiated from the 15 M{sub o-dot} progenitor.

Verification of NUREC Code Transient Calculation Capability Using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem

International Nuclear Information System (INIS)

Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon

2006-01-01

In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes

Aspects of cell calculations in deterministic reactor core analysis

International Nuclear Information System (INIS)

Varvayanni, M.; Savva, P.; Catsaros, N.

2011-01-01

Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

SASSYS LMFBR systems code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

1983-01-01

The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

International Nuclear Information System (INIS)

T-M Sembiring; S-Pinem; P-H Liem

2015-01-01

The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

Energy Technology Data Exchange (ETDEWEB)

Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

2016-12-15

Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

Analysis of a 12-Finger Rod Drop using RETRAN/MASTER Code System for APR1400

International Nuclear Information System (INIS)

Yu, Keuk Jong; You, Sung Chang; Kim, Han Gon

2009-01-01

The Optimized Power Reactor 1000 (OPR1000) has 4-finger and 12-finger Control Element Assemblies (CEAs). When the 12-finger CEA is dropped, Core Protection Calculator System (CPCS) shuts down the reactor to prevent fuel damage that could occur from the sudden reactor power peaking. By contrast, the improved CPCS of Advanced Power Reactor 1400 (APR1400), which has systems similar to those of the OPR1000, decreases reactor power rapidly using its Reactor Power Cutback System (RPCS) to avoid unwanted reactor trips caused by the CPCS during a 12- finger CEA drop event. RETRAN is a best-estimate code for transient analysis of Non-LOCA. The RETRAN control logic, which includes the function of reducing reactor power during a 12-Finger CEA drop, has been developed for the APR1400. A MATRAN program has also been developed. MATRAN is the interface program for realtime processing to connect RETRAN with MASTER code which is a nuclear analysis and design code. MATRAN supplies adequate feedback reactivities from the MASTER code to RETRAN code. The purpose of this study is to analyze the behavior of a nuclear reactor core and its primary system using conventional RETRAN analysis procedure and MATRAN program analysis procedure during a 12- finger CEA drop. In addition, the axial power distribution and Axial Shape Index (ASI) are produced by the MATRAN program and they are confirmed as within operation limits

Development of steady thermal-hydraulic analysis code for China advanced research reactor

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

2006-01-01

A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

Benchmark analysis of SPERT-IV reactor with Monte Carlo code MVP

International Nuclear Information System (INIS)

Motalab, M.A.; Mahmood, M.S.; Khan, M.J.H.; Badrun, N.H.; Lyric, Z.I.; Altaf, M.H.

2014-01-01

Highlights: • MVP was used for SPERT-IV core modeling. • Neutronics analysis of SPERT-IV reactor was performed. • Calculation performed to estimate critical rod height, excess reactivity. • Neutron flux, time integrated neutron flux and Cd-ratio also calculated. • Calculated values agree with experimental data. - Abstract: The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of K eff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results

Achieving 95% probability level using best estimate codes and the code scaling, applicability and uncertainty (CSAU) [Code Scaling, Applicability and Uncertainty] methodology

International Nuclear Information System (INIS)

Wilson, G.E.; Boyack, B.E.; Duffey, R.B.; Griffith, P.; Katsma, K.R.; Lellouche, G.S.; Rohatgi, U.S.; Wulff, W.; Zuber, N.

1988-01-01

Issue of a revised rule for loss of coolant accident/emergency core cooling system (LOCA/ECCS) analysis of light water reactors will allow the use of best estimate (BE) computer codes in safety analysis, with uncertainty analysis. This paper describes a systematic methodology, CSAU (Code Scaling, Applicability and Uncertainty), which will provide uncertainty bounds in a cost effective, auditable, rational and practical manner. 8 figs., 2 tabs

Analysis of PHEBUS FPT1 test with IMPACT/SAMPSON code

International Nuclear Information System (INIS)

Terada, Masafumi; Ikeda, Takashi; Naitoh, Masanori

2003-01-01

IMPACT is a simulation software developed at the Nuclear Power Engineering Corporation, which includes the severe accident analysis code, SAMPSON. SAMPSON consists of twelve modules and is capable of simulating hypothesized severe accidents in LWR. Phebus-FPT1 test, which was selected as the International Standard Problem-46, was analyzed with SAMPSON for the verification of the code. The Phebus-FPT1 test was an integral in-pile experiment for studying mainly degradation of fuel bundle and subsequent FP behavior under a LWR severe accident condition, using irradiated fuel as a source of real FP. The following analyses of the Phebus-FPT1 test, which are also the subjects of the ISP-46, were performed: (1) In-core thermal hydraulics, core degradation and FP release from the fuel, (2) FP gas and aerosol transport in the primary circuit, (3) Thermal hydraulics and FP aerosol physics in the containment and (4) Iodine chemistry in the containment. The analysis results of the thermal hydraulics and core degradation showed good agreement with experimental data, except shroud temperatures which were higher than the experiment. The difference may be due to insufficient modeling of the gap closure in the shroud. FP release from fuel, FP transport rate in the primary circuit, FP aerosol physics and iodine chemistry in the containment were also well predicted. Through the analyses, the modules of SAMPSON used were proved to be capable for evaluating thermal hydraulics and FP behaviors under LWR severe accident conditions

Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

International Nuclear Information System (INIS)

Lajtman, S.; Feretic, D.; Debrecin, N.

1989-01-01

This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Directory of Open Access Journals (Sweden)

Siniša Šadek

2010-01-01

Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

International Nuclear Information System (INIS)

Chvetsov, I.; Kouznetsov, I.; Volkov, A.

2000-01-01

GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

Coupling of the computational fluid dynamics code ANSYS CFX with the 3D neutron kinetic core model DYN3D

International Nuclear Information System (INIS)

Kliem, S.; Grahn, A.; Rohde, U.; Schuetze, J.; Frank, Th.

2010-01-01

The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in

BN-600 MOX Core Benchmark Analysis. Results from Phases 4 and 6 of a Coordinated Research Project on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects

International Nuclear Information System (INIS)

2013-12-01

For those Member States that have or have had significant fast reactor development programmes, it is of utmost importance that they have validated up to date codes and methods for fast reactor physics analysis in support of R and D and core design activities in the area of actinide utilization and incineration. In particular, some Member States have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems, it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a coordinated research project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. The CRP started in November 1999 and, at the first meeting, the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP, investigating different nuclear data and levels of approximation in the calculation of safety related reactivity effects and their influence on uncertainties in transient analysis prediction. In an additional phase of the benchmark studies, experimental data were used for the verification and validation of nuclear data libraries and methods in support of the previous three phases. The results of phases 1, 2, 3 and 5 of the CRP are reported in IAEA-TECDOC-1623, BN-600 Hybrid Core Benchmark Analyses, Results from a Coordinated Research Project on Updated Codes and Methods to Reduce the

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

International Nuclear Information System (INIS)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

2001-01-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

Energy Technology Data Exchange (ETDEWEB)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

2001-07-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

International Nuclear Information System (INIS)

Matijevic, M.; Jecmenica, R.; Pevec, D.; Trontl, K.

2012-01-01

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals of the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working enviroment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 code system, with the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analysed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be

A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

International Nuclear Information System (INIS)

Kim, Hong Chul

2005-02-01

In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

Extension of the analytic nodal diffusion solver ANDES to triangular-Z geometry and coupling with COBRA-IIIc for hexagonal core analysis

International Nuclear Information System (INIS)

Lozano, Juan-Andres; Jimenez, Javier; Garcia-Herranz, Nuria; Aragones, Jose-Maria

2010-01-01

In this paper the extension of the multigroup nodal diffusion code ANDES, based on the Analytic Coarse Mesh Finite Difference (ACMFD) method, from Cartesian to hexagonal geometry is presented, as well as its coupling with the thermal-hydraulic (TH) code COBRA-IIIc for hexagonal core analysis. In extending the ACMFD method to hexagonal assemblies, triangular-Z nodes are used. In the radial plane, a direct transverse integration procedure is applied along the three directions that are orthogonal to the triangle interfaces. The triangular nodalization avoids the singularities, that appear when applying transverse integration to hexagonal nodes, and allows the advantage of the mesh subdivision capabilities implicit within that geometry. As for the thermal-hydraulics, the extension of the coupling scheme to hexagonal geometry has been performed with the capability to model the core using either assembly-wise channels (hexagonal mesh) or a higher refinement with six channels per fuel assembly (triangular mesh). Achieving this level of TH mesh refinement with COBRA-IIIc code provides a better estimation of the in-core 3D flow distribution, improving the TH core modelling. The neutronics and thermal-hydraulics coupled code, ANDES/COBRA-IIIc, previously verified in Cartesian geometry core analysis, can also be applied now to full three-dimensional VVER core problems, as well as to other thermal and fast hexagonal core designs. Verification results are provided, corresponding to the different cases of the OECD/NEA-NSC VVER-1000 Coolant Transient Benchmarks.

Comparative analysis of the results obtained by computer code ASTEC V2 and RELAP 5.3.2 for small leak ID 80 for VVER 1000

International Nuclear Information System (INIS)

Atanasova, B.; Grudev, P.

2011-01-01

The purpose of this report is to present the results obtained by simulation and subsequent analysis of emergency mode for small leak with ID 80 for WWER 1000/B320 - Kozloduy NPP Units 5 and 6. Calculations were performed with the ASTEC v2 computer code used for calculation of severe accident, which was designed by French and German groups - IRSN and GRS. Integral RELAP5 computer code is used as a reference for comparison of results. The analyzes are focused on the processes occurring in reactor internals phase of emergency mode with significant core damage. The main thermohydraulic parameters, start of reactor core degradation and subsequent fuel relocalization till reactor vessel failure are evaluated in the analysis. RELAP5 computer code is used as a reference code to compare the results obtained till early core degradation that occurs after core stripping and excising of fuel temperature above 1200 0 C

Severe core damage experiments and analysis for CANDU applications

International Nuclear Information System (INIS)

Mathew, P.M.; White, A.J.; Snell, V.G.; Bonechi, M.

2003-01-01

AECL uses the MAAP CANDU code to calculate the progression of a severe core damage accident in a CANDU reactor to support Level 2 Probabilistic Safety Assessment and Severe Accident Management activities. Experimental data are required to ensure that the core damage models used in MAAP CANDU code are adequate. In SMiRT 16, details of single channel experiments were presented to elucidate the mechanisms of core debris formation. This paper presents the progress made in severe core damage experiments since then using single channels in an inert atmosphere and results of the model development work to support the experiments. The core disassembly experiments are conducted with one-fifth scale channels made of Zr-2.5wt%Nb containing twelve simulated fuel bundles in an inert atmosphere. The reference fuel channel geometry consists of a pressure tube/calandria tube composite, with the pressure tube ballooned into circumferential contact with the calandria tube. Experimental results from single channel tests showed the development of time-dependent sag when the reference channel temperature exceeded 850 degC. The test results also showed significant strain localization in the gap at the bundle junctions along the bottom side of the channel, thus suggesting creep to be the main deformation mechanism for debris formation. An ABAQUS finite element model using two-dimensional beam elements with circular cross-section was developed to explain the experimental findings. A comparison of the calculated central sag (at mid-span), the axial displacement at the free end of the channel and the post-test sag profile showed good agreement with the experiments, when strain localization was included in the model, suggesting such a simple modelling approach would be adequate to explain the test findings. The results of the tests are important not only in the context of the validation of the analytical tools and models adopted by AECL for the severe accident analysis of CANDU reactors but

Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 2. BFS-62 analysis results

International Nuclear Information System (INIS)

Mantourov, Guennadi

2001-11-01

This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. Data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS-62 and ZPPR JUPIER series fast reactor cores, applying JNC core calculation code CITATION-FBR. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the First Flight Collision Probability method and subgroup approach. Especially, two converting programs were written to transmit the prepared effective cross sections to JNC standard PDS files to let then the CITATION code be applied for 3-D HEXZ neutronics calculations of the investigated cores. The effects of nuclear data library have been studied by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for 4 BFS-62 cores as well as for 3 ZPPR JUPITER experiment series cores ZPPR-9, ZPPR-13A and ZPPR-17A are presented. The comparison results for reaction rates distributions for 2 BFS-62 uranium loaded cores are included too. The calculated correction factors applied in all cases were less than 1.0%. The estimated uncertainty in k-effective C values caused by possible errors in calculation of the applied corrections is about 0.3% in case of BFS-62 and ZPPR MOX cores, and is about 0.2% for BFS-62 uranium-loaded cores. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC's calculation route for k-effective results is about 0.3% for ZPPR and BFS-62 cores with plutonium. As for BFS uranium-loaded cores (BFS-62-1 and BFS-62-2) the nuclear data library effect is about 0.1%. Next the sensitivity analysis was applied. It shown that the main contributors to the nuclear data library effect

The symbol coding language for the BUTs processor of in-core reactor control systems

International Nuclear Information System (INIS)

Vorob'ev, D.M.; Golovanov, M.N.; Levin, G.L.; Parfenova, T.K.; Filatov, V.P.

1978-01-01

A symbolic coding language is described; it has been developed for automation of making up programs for in-core control systems. The systems use the ideology of the CAMAC-VECTOR system and include the BUTs-20 processor. The symbolic coding language has been developed as a programming language of the ASSEMBLER type. Operators of instructions and pseudo-instructions, the rules of reading in the text of the source program, and operator record formats are considered

Analysis of Homogeneous BFS-73-1 MA Benchmark Core

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

2007-06-15

Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.

Analysis of Homogeneous BFS-73-1 MA Benchmark Core

International Nuclear Information System (INIS)

Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

2007-06-01

Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty

An approach to model reactor core nodalization for deterministic safety analysis

International Nuclear Information System (INIS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH 1.6 , stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D ® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M

An approach to model reactor core nodalization for deterministic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

2016-01-22

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis